ML20042F509

Proposed Tech Specs Re long-term Defueling

ML20042F509
Person / Time
Site:	Rancho Seco
Issue date:	04/30/1990
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20042F507	List: ML20042F506 ML20042F509 ML20043A086 ML20043A088
References
NUDOCS 9005090007
Download: ML20042F509 (77)

Text

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[,' ~ RANCH 0 SECO UNIT 1- ~ R 'LONG TERM DEFUELED c-TECHNICAL SPECIFICATIONS l LINDEX + DEFINITIONS SECTION PAGE D1.0 DEFINITIONS...................... D1-1 Dl.1-ACTION:........................ D1-1 D1.2 LONG TERM DEFUELED MODE................. Dl-1 01.3 OPERABLE - OPERABILITY,................ D1 Dl.4 -INSTRUMENTATION SURVEILLANCE........_..... D1-1 D1.4.1.- CHANNEL TEST........... _.......... D1 D1.4.2 INSTRUMENT CHANNEL CHECK............... D1-1 'D1.4.3 INSTRUMENT CHANNEL CALIBRATION............ Dl-1 DI.4.4 FUNCTIONAL TEST........... D1-2 j D1. 5' SURVEILLANCE INTERVALS................ D1-2 D1.6 PROCESS CONTROL PROGRAM (PCP) Dl-2 1 D1.7 -0FFSITE DOSE CALCULATION MANUAL (00CM) Dl-2 101.8 MEMBER (S) 0F THE PUBLIC................ D1-2 D1.9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) l MANUAL........................ 01-3 i D1.10 NUCLEAR SAFETY...., 01-3 l ^l l \\ t . LIMITING' CONDITIONS AND SURVEILLANCE-REOUIREMENTS J SECTION PAGE l s^ D3.0-GENERAL LIMITING CONDITIONS..-............ D3/4-1 D4.0 GENERAL SURVEILLANCE REQUIREMENTS........... D3/4-1 q l D3/4.1 SPENT FUEL P0OL......-.............. 03/4-2 3 D3/4.2 SPENT FUEL POOL TEMPERATURE.............. D3/4-3 1 D3/4.3 FUEL STORAGE BUILDING HANDLING LOAD LIMITS...... 03/4-4 03/4 4 SPENT FUEL STORAGE RADIATION MONITOR,........ 03/4-5 1 03/4.5 SPENT FUEL P0OL HATER ~ CHEMISTRY............ D3/4-6 03/4.6 ~ LIQUID HOLD-UP TANKS................. 03/4-7

D3/4.7 SEALED SOURCE LEAK TESTING..............

D3/4-8 1 Proposed Amendment No. 182, Revision 1 DI-1 1 1

7 q l RANCHO SECO UNIT 1 'LONG TERM DEFUELED TECHNICAL SPECIFICATIONS-l T INDEX. 1 DESIGN FEATURES' g SECTI0ff PAGE 05.1 SITE'......................... D5-1 05.1.1 EMERGENCY PLANNING ZONE (EPZ). DS-1 l-{ D5.1.2 SITE BOUNDARY FOR GASEOUS EFFLUENT,......... 05-1 f 05.1.3 BOUNDARY FOR LIQUID EFFLUENT............. D5 'l j 05.2~ . SPENT-FUEL STORAGE FACILITIES............. D5-5 05.2.1 SPENT FUEL STORAGE RACKS AND FAILED FUEL . STORAGE CONTAINER RACK................ D5-5 5.2.2-SPENT FUEL POOL AND STORAGE RACK DESIGN........ D5-5 ADMINISTRATIVE CONTROLS j SECTION' EAGE D6.1-RESPONSIBILITY,................... 06-1 D6.2-ORGANIZATION..................... D6-1 -D6.3 FACILITY STAFF QUALIFICATIONS............. D6-3 .06.4 ~ TRAINING..., 06-3 06.5 REVIEH AND AUDIT................... D6-3 1D6.5.1 NUCLEAR SAFETY REVIEH AND AUDIT COMMITTEE (NSRAC). D6-3 D6.5.2- -NOT USED '........-.............. D6-5 l ' 6 '. 5. 3 TECHNICAL REVIEH AND CONTROL.,............. D6-5 D6.5.4 AUDITS-.....-................... D6-6 D6.6 -NOT USED...................... D6-7' D6.7~ .-NOT USED....................... D6-7 D6.8 PROCEDURES, PLANS, MANUALS, AND PROGRAMS....... D6-8 D6.9 REPORTING REQUIREMENTS................ D6-10 D6.9.1L ANNUAL RADIOLOGICAL REPORTS.............. D6-10 t D6.9.21 SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT,... D6-11 .i D6.9.3 / ANNUAL REPORT..................... D6-11 D6.9.4 LICENSEE EVENT REPORT................. D6-12 .D6.9.5 -NOT USED...................... 06-12 D6.9.6 ENVIRONMENTAL REPORTS................. 06-12 06.10-RECORD RETENTION..,,............... D6-13 D6.11 RADIATION PROTECTION PROGRAM............. D6-14 'D6.12 HIGH RADIATION AREA.................. D6-15 D6.13 PROCESS CONTROL PROGRAM (PCP). D6-15 D6.14 0FFSITE DOSE CALCULATION AND RADIOLOGICAL . ENVIRONMENTAL MONITORING PROGRAM MANUALS....... D6-16 Proposed Amendment No.182, Revision 1 i DI-2

TC 1 . RANCH 0'SECO UNIT 1-LONG TERM DEFUELED LTECHNICAL SPECIFICATIONS l'e . LIST.0F: TABLES IdbleL PAga D1. 5-.1 TIME PERIODS-01-3 1 D3.5-1 SPENT' FUEL POOL HATER CHEMISTRY D3/4-6' D6.2-1 MINIMUM SHIFT CREH' REQUIREMENTS D6 L LIST OF FIGURES - Figure-Egge F - D5.1 EMERGENCY PLANNING ZONE-(EPZ) 05-2' l] 05.1-2 . SITE BOUNDARY FOR GASEQUS EFFLUENT DS-3 'l D5.1-3 BOUNDARY FOR LIQUID-EFFLUENT DS-4 l.j i ,3 i i 1 -i 'I j i .i 5 Proposed Amendment No. 182, Revision 1 DI-3 !)

y I i RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS D1;0 DEFINITIQRS. The-following terms are defined for uniform interpretation of these technical specifications. 'D1.1 ACTION J ACTION including time requirements shall be that part of a specification which prescribes remedial measures required under designated conditions. 'D1.2 LONG TERM DEFUELED MODE The plant is in~ a LONG TERM DEFUELED MODE when no fuel is in the Reactor Building and the fuel remaining on site is stored in the spent fuel pool. The LONG TERM DEFUELED MODE reficcts the District's intent-to safely store spent fuel but not operate the reactor or move fuel into the. Reactor Building. D1.3 OPERABLE - OPERABILITY A component or system is OPERABLE when it is capable of performing its intended function-within the required. range. The component or system shall be considered to have this capability when: (1) it satisfies the Limiting ' Conditions defined in Specification D3, (2) it-has been tested periodically-in-accordance with Specification D4 and has met its performance requirements, (3) the system has available its normal source of power, and (4) its required auxiliaries are capable of performing their intended function. D1.4-INSTRUMENTATION SURVEILLANCE D1.4.1 CHANNEL TEST . A CHANNEL TEST is the injection of an internal or external test signal into the channel to verify its proper response, including alarm and/or trip initiating action, where applicable. D1.4.2 INSTRUMENT CHANNEL CHECK An INSTRUMENT CHANNEL CHECK is a' verification of acceptable instrument performance by observation of its behavior and/or state. This verification includes comparison of output and/or state of independent channels measuring the same variable if there is redundant indication. D1.4.3 INSTRUMENT CHANNEL CALIBRATION An INSTRUMENT CHANNEL CALIBRATION is a test, and adjustment (if necessary), to 1 establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values. INSTRUMENT CHANNEL CALIBRATION shall encompass the entire channel, including equipment actuation, alarm, or trip, whichever are applicable, and shall be deemed to include the CHANNEL TEST. Proposed ' Amendment No. 182, Revision 1 01-1

I RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS D1.4.4 FUNCTIONAL TEST A FUNCTIONAL TEST shall.be.the' determination or verification or the capability of a system or component to. meet specified requirements by subjecting the systein or component to a set of physical or operating conditions. D1.5 SURVEILLANCE-INTERVALS f - The SURVEILLANCE INTERVAL may be extended to a maximum of +2St. to accommodate operations scheduling. The frequency notation specified for the performance of Surveillance Requirements shall correspond to the SURVEILLANCE INTERVALS' defined in Table D1.5-1. D1.6 PROCESS CONTROL PROGRAM (PCP) The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling analyses, test, and determinations to be made to ensure that

i processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, state

' regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste. Dl.7 0FFSITE DOSE CALCULATION MANUAL (ODCM) The 0FFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive . gaseous and liquid effluents and used in the calculations of gaseous and -liquid effluent monitoring Alarm / Trip Setpoints. The ODCM shall also contain the Radioactive Effluent Controls Program required by Specification D6.8.3a and descriptions of the information that should be included in the Semiannual . Radioactive Effluent Release Report required by Specification D6.9.2. D1.8 MEMBER (S)-OF THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include individuals who by virtue of.their occupational-status have no formal association with the plant..This category shall include nonemployees of the licensee who are permitted to use portions of. the site for recreational, occupational, or other purposes not associated -with-plant functions. This category shall n0_t include nonemployees such as vending machine servicemen or postmen who, as part of their. #ormal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. Proposed Amendment No. 182, Revision 1 D1-2

RANCHO SECO UNIT 1 t LONG TERM DEFUELED TECHNICAL SPECIFICATIONS DI.9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL The RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) MANUAL shall contain

a. description of the Rancho Seco radiological environmental monitoring program.

The REMP MANUAL shall also contain the REMP requirements of . Specification D6.8.3b, a description of the environmental samples to be collected, sample locations, sampling frequencies, sample analysis criteria, and a description of the information to be included in the Annual Radiological Environmental Operating Report as requ' red by Specification D6.9.1.3. i D1.10 NUCLEAR SAFETY a NUCLEAR SAFETY shall refer to those systems, components, and administrative controls that have or may have an effect on the health and safety of the public. TABLE D1.5-l* SURVEILLANCE INTERVALS Freauency Notation Qpfinition i = DAILY D At least once per 24 hours. HEEKLY H At least once per 7 days. MONTHLY M At least once per 31 days. QUARTERLY Q At least once per 92 days, l ANNUALLY A At least once per 12 months. 6 MONTHS 6M At least once per 6 months. 18 MONTHS 18M At least once per 18 months. l See Specification Dl.5. 1 i Proposed Amendment No.182, Revision 1 Dl-3 l.. l

~_ RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS D3.0 GENERAL LIMITING CONDITIONS LIMITING CONDITION l t D3.0.1 Compliance with the Limiting Conditions contained in the Section D3 Specifications is required during the LONG TERM DEFUELED MODE, except that upon failure to meet the Limiting Conditions, the associated ACTION including time requirements shall be met. D3.0.2 Noncompliance withla specification shall exist when the requirements of the Limiting. Condition and associated ACTION including time requirements are not met within the specified time intervals. If the Limiting Condition is ~ restored prior to expiration of the specified time intervals, completion of the ACTION ~ including time requirements is not required. l SURVEILLANCE REOUIREMENTS I D4.0 GENERAL SURVEILLANCE REOUIREMENTS D4.0.1. Surveillance Requirements shall be met as specified for individual Limiting Conditions unless otherwise stated in an individual Surveillance Requirement. D4.0.2 Failure to perform a Surveillance Requirement within the specified SURVEILLANCE, INTERVAL:with a maximum allowable extension not to exceed 25% of -the specified SURVEILLANCE INTERVAL (reference Specification D1.5) shall 3 constitute noncompliance with the OPERABILITY requirements for a Limiting Condition..The time clock for an ACTION starts at the time it is identified that'a Surveillance Requirement has not been met. An ACTION including time requirements may be delayed for up to 24 hours to permit the completion of the survoillance when the allowable outage time limits of the ACTION including time requirements are less than 24 hours. Exceptions to these requirements are stated in the individual specifications. l ? s l Proposed l.: Amendment No. 182, Revision 1 D3/4-1 l

a RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS 'D3/4.1 SPENT FUEL POOL LIMITING CONDITION 3 D3.' l At least 37 feet of water shall be maintained in the spent fuel pool. D3.1.1 The water level in the spent fuel pool may be less than.37 feet if the dose rate at the pool surface, from irradiated core components seated in the spent fuel pool storage racks, is 2.5 mrem /hr or less. APPLICABILITY: Whenever irradiated core components are in the spent fuel pool. ACILON: With a requirement (s) of this Limiting Condition not satisfied, suspend movement of irradiated core components and crane operations with loads over stored spent fuel, and restore the spent fuel pool to within its Limiting Condition within 4 hours. SURVEILLANCE REOUIREMENTS 04.1.1 DAILY verify the level in the spent fuel pool is at least at the minimum required level; if less than 37 feet, perform a radiation survey at the pool surface level.to ensure 2.5 mrem /hr or less. D4.1.2 At least once every 18 MONTHS perform an INSTRUMENT CHANNEL CALIBRATION on the spent fuel pool level alarm switches. Bases When the spent fuel pool water level is maintained at a minimum of 37 feet, a minimum of 23 feet of water shielding over stored fuel assemblies is assured. This level limits radiation at the surface of the water to less than 2.5 mrem /hr. A water level of less than 37 feet is allowed as long as the dose rate at the surface of the spent fuel pool, from irradiated core components seated in their storage locations in the Spent Fuel Pool, is 12.5 mrem /hr. Proposed Amendment No. 182, Revision 1 D3/4-2

7; -. RANCHO SECO UNIT'1' LONG TERM DEFUELED 1 o TECHNICAL SPECIFICATIONS L D3/4.2 SPENT-FUEL POOL TEMPERATURE LIMITING CONDITION 03.2 Maintain the spent' fuel ' pool bulk coolant temperature'below 140*F. Whenever irradi' ted fuel asserablies are in the spent fuel pool APPLICABILITY:. a ACILOR: - Hith the-Spent Feel Pool Cooling System not maintaining spent fuel pool bulk ) coolant temperature below 140'F, place an alternate cooling method in r service. Also, prior to exceeding a bulk coolant temperature of 180*F, place an alternate cooling method in service to supplement the Spent Fuel Pool Cooling System. SURVEILLANCE REOUIREMENTS D4.2 DAILY verify the spent fuel pool bulk coolant temperature is <140'F. Bases This specification provides a method to ensure that the spent' fuel pool bulk - coolant temperature does not reach the boiling point.. The use of an alternate, 1 cooling method provides assurance that boiling will not occur in the spen. 3: fuel pool. s t T ' Proposed 1 Amendment No. 182, Revision 1 D3/4-3 -i ' i. -5

a fi RANCHO SECO UNIT 1 e LONG TERM DEFUELED TECHNICAL-SPECIFICATIONS Q3/4.3 FUEL STORAGE BUILDING HANDLING LOAD LIMITS LIMITING CONDITION c D3.3 Loads'in excess of 1700 pounds shall be prohibited from traveling over irradiated fuel assemblies in the spent fuel pool. AEEk1CAQ1LIJX: Whenever irradiated fuel assemblies are in the spent fuel pool ACTION: Hith the requirements of the above specification not met, place the Fuel Storage Building fuel handling bridge and overhead crane in a condition such that loads in excess of 1700 pounds are not over irradiated fuel assemblies. SURVEILLANCE REOUIREMENTS 04.3.1 Perform a dead weight load test at the rated load on the crane (s) to be.used within 7 days of handling fuel assemblies or moving loads over irradiated fuel assemblies stored in the spent fuel pool. D4.3.2 Complete a FUNCTIONAL TEST of the Fuel Storage Building fuel handling bridge interlocks within 7 days prior to fuel handling. Bases The restriction on movement of loads greater than 1700 pounds over fuel assemblies in the spent fuel pool ensures that in the event a load is dropped: _ (1) the activity release will be limited to that contained in a single fuel assembly, and (i.) distortion of fuel in the spent fuel pool storage racks will not result in a critical array. These assumptions'are -consistent with the accident analysis for a dropped fuel assembly. The rated load of the Fuel Storage Eu11 ding-fuel handling bridge is 2900 pounds. The rated load of the overhead crane in the Fuel' Storage Building is 3000 pounds, h l \\ Proposed Amendment No. 182, Revision 1 l D3/4-4 1

i RANCHO SECO UNIT 1 LONG TERM DEFUELED o TECHNICAL SPECIFICATIONS D3/4'.4 SPENT FUEL STORAGE RADIATION MONITOR i LIMITING CONDITION D3.4 Radiation-levels in the spent fuel storage area shall be monitored by a fixed radiation monitor. j i ' APPLICABILITY:' Whenever irradiated fuel assemblies are in the spent fuel pool l l ACTION: l .Hith fixed radiation monitoring equipment inoperable, suspend fuel handling operations,. place-fuel in a safe condition, and attempt to return a fixed radiation monitor'to OPERABLE status, i I If a fixed radiation monitor is not returned to OPERABLE status within 7 days, prepare and submit a report within an additional 30 days to the NRC, Region V documenting the condition along with a plan to return a fixed radiation monitor to OPERABLE' status. Also, fuel handling operations may resume as long i as portable radiation survey instrumentation, having the appropriate ranges and sensitivity to protect individuals involved in fuel hWiling operations, is used uritil a fixed radiation monitor is returned to OPENABLE status. SURVEILLANCE REOUIREMENTS D4.4.1. DAILY during fuel handling operations, or WEEKLY otherwise, perform an INSTRUMENT CHANNEL CHECK. D4.4.2 MONTHLY. perform an instrument CHANNEL TEST. ] D4.4.3 QUARTERLY perform an INSTRUMENT CHANNEL CALIBRATION. Bases i The' use of radiation monitoring equipment ensures that early notification of excessively high radiation levels in the Spent Fuel Building is provided to protect individuals involved in fuel handling operations. 1 l i l Proposed L Amendment No. 1B2, Revision 1 D3/4-5

i c RANCHO SEC0' UNIT 1-LONG TERM DEFUELED TECHNICAL SPECIFICATIONS D3/4.5 SPENT FUEL POOL HATER CHEMISTRY LIMITING CONDITION D3.5 The spent fuel pool water chemistry shall be maintained within the limits specified in Table D3.5-1. APPLICABILITY:' AT'ALL TIMES AC110H: 'Hith a water chemistry limit exceeded, initiate action within 72 hours to restore the water chemistry to within limits and conduct an evaluation to j_ ' determine the caust. SURVEILLANCE REOUIREME.NTS D4.5 The spent fuel pool water chemistry shall be determined to be within the' limits by analysis for the parameters and at the frequencies specified in Table D3.5-1. Bases The maintenance of spent fuel pool water chemistry ensures.that degradation of irradiated core components, the spent fuel pool liner, and the fuel storage racks is minimized. l TABLE D3.5-1 SPENT FUEL POOL HATER CHEMISTRY Parameter Units Limit Analysis Freauency Chloride ppm 50.15 MONTHLY Fluoride ppm 50.15 MONTHLY v l. ' Proposed L Amendment No. 182, Revision 1 D3/4-6

RANCHO SECO UNIl 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS' D1/4.6 LIOUID HOLD-UP TANKS ' LIMITING CONDITION .D3.6 The quantity'of radioactive material contained in each of the following ) tanks shall be limited to 110 Curies, excluding Tritium and dissolved or entrained noble gases: a. A or B Regenerant Hold-up Tanks { b. Demineralized Reactor Coolant Storage Tank

c. : Miscellaneous Hater Hold-up Tank d.

Borated Water Storage Tank e. Outsic'e Temporary Tanks APPLICABILJll:. AT ALL TIMES ' ACILOS: Hith the quantity.of radioactive material in any of the listed tanks exceeding the above limit,-immediately suspend all additions of radioactive material to the tank, and initiate actions to reduce the-tank contents to within the - limit. Reduce the-tank's contents to within the limit within the next 72 hours and describe the events leading to this condition.in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 06.9.2. SURVEILLANCE REOUIREMENTS 04.6 The quantity of radioactive material contained in each tank listed in Specification D3.6 shall be determined to be within the specified limit by analyzing a representative sample of the tank's contents within 7 days following addition of radioactive materials to the tank. Bases Restricting the quantity of radioactive material contained in the specified outdoor tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentration at the nearest potable water supply and the nearest surface water supply in an unrestricted area would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2. The limit applies to.enh tank individually. Tanks included in this specification are those outdoor tanks that contain radioactive liquid, are not surrounded by liners, dikes, or walls capable of holding theflank contents, or do not have tank overflows and surrounding area drains connected to a liquid radwaste treatment system. 2 Proposed Amendment No. 182. Revision 1 D3/4-7 4

p: i ? RANCHO SECO UNIT 1 g LONG TERM DEFUELED- [ TECHNICAL SPECIFICATIONS l D3/4.7 SEALED SOURCE LEAK TESTING LIMITING CONDITION i -03.7-Each sealed source containing radioactive material either in excess of - 100 microcuries of beta'and/or gamma emitting material or 10 microcuries of - alpha and/or neutron emitting material shall be periodically verified to have I 0.005 microcuries of removable contamination on its surface. 1 APPLICABILITY: AT ALL TIMES, except that sealed sources containing Tritium, licensed material with a half-life of 130 days, licensed material in gaseous form,1100 microcuries. of beta and/or gamma emitting material, and.110 microcuries of alpha and/or neutron emitting material are exempt from the requirements of this Specification. ACTION: 1 Hith a sealed source having removable contamination in excess of an above limit, _immediately withdraw the sealed source from use and either: a. Decontaminate and repair the sealed source, or b. Dispose of the sealed source in accordance with NRC Regulations. SURVEILLANCE RE0UIREMENTS D4.7.1 Test Reauirements Each sealed source, meeting -the APPLICABILITY criteria above, shall be tested for leakage and/or. contamination by: i a. The Licensee or other person specifically authorized by the NRC, or s b. An' Agreement State. The test method -shall have a detection sensitivity of at least 0.005 microcuries per test sample. D4.7.2 Test Freauencies These sealed sources, excluding sealed sources that are being stored and not used and the startup sources and fission detectors previously subjected to core flux, shall be tested once every 6 MONTHS. J Sealed sources that are being stored and not used, and the startup sources and fission detectors previously subjected to core flux shall be testeri prior to use or transfer to another licensee unless they have been tested within the previous 6 months. Sealed sources that are received but do not have a certificate which indicates a test for leakage was performed during the 6 months before the transfer shall be tested prior to being placed into service. Proposed -Amendment No. 182, Revision 1 D3/4-8

e-V RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS D3/4.7 SEALED SOURCE-LEAK TESTING (continued) Bases i .This Specification meets the requirements of 10 CFR-39.35, Leak Testing of 2 -Sealed Sources. The limitations on removable contamination for sources containing plutonium is specified in 10 CFR 70.39(c). The performance of-i l. sealed source leak-testing by a person specifically authorized by the NRC.or

j an Agreement State is permitted by 10 CFR 39.35(b).

>l -l 1 Sealed sources are classified into three groups according to their use, with' surveillance requirements commensurate with the probability of damage to a source in that group. Sealed sources which are continuously enclosed within a i shielded mechanism (i.e., sealed sources within radiation monitoring or boron j ' measuring devices) are considered to be stored and need not be tested unless j they are removed from the shielded mechanism. i i i l 1 l -i Proposed - Amendment No.182, Revision 1 D3/4-9

7;. e, T RANCHO SEC0' UNIT 1-LONG TERM DEFUELED TECHNICAL SPECIFICATIONS D5.0 DESIGN FEATURES D5.1 SITE The Rancho Seco site is located on the 2,480 acres owned by Sacramento Municipal Utility District, 26 miles north-northeast of Stockton and 25 miles southeast of the City of, Sacramento, California. USAR Figure 1.1-2 shows~the plan of the site. The Emergency Planning Zone for the LONG TERM DEFUELED H0DE is the site industrial area. -D5.1.1 Emeroency Plannina Zone The Emergency Planning Zone is shown in Figure 05.1-1. It is the same as the Industrial Area Boundary. D5.1.2 Site boundary For Gaseous Effluent The' Site Boundary for Gaseous Effluent for 10 CFR 20 compliance and for meeting 10 CFR 50, Appendix I guidelines is shown in Figure DS.1-2. D5.1.3 Boundarv For Liauid Effluent The boundary for Liquid Effluent-for 10 CFR 20 Compliance is shown in Figure D5.1-3. The dose accountability ~ points for meeting the 10 CFR 50, Appendix I guidelines remains at the A and B Regenerant Hold-Up Tanks. Concentration accountability points for determining 10 CFR 20, Appendix B compliance remains at the Retention Basins. P t l l-Proposed i Amendment No.182, Revision 1 05-1

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1 RANCHO SECO UNIT ~l' LONG TERM DEFUELEO TECHNICAL SPECIFICATIONS- [ FIGURE 05.1-3 BOUNDARY FOR LIQUID EFFLUENT-i t i t 1 'Y .Q 1 au m Y y ~ l-l ..u, s.....,. -/ i '.Il.B; ~ sp. g=p,,.. q j esino.u s.ani.

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fi E p n' RANCHO SECO UNIT 1 2 LONG TERM DEFUELED-TECHNICAL SPECIFICATIONS 'D5.2' SPENT FUEL STORAGE FACILITIES i

Q5 2.1 Soent Fuel Storage Racks and Failed Fuel Storage Container Rack 1

Irradiated or failed fuel shall be stored in the stainless steel lined spent r fuel pool prior to offsite shipment..The spent fuel pool is sized to accommodate 1080 fuel assemblies, plus four assemblies in failed fuel containers. The pool has.the capability of storing fuel assemblies in eleven freestanding stainless steel rack modules and four failed fuel assemblies in a special rack module, All assemblies are'on nominal 10.5 inch centers in both directions. This spacing with the neutron absorber material is sufficient to maintain. Keff_ <0.95 when flooded with unborated water, based on a fuel enrichment of 4.0 weight percent. D5.2.2 Soent Fuel Pool-and Storaae Rask Desian The spent fuel pool and all storage racks are designed for the plant Design Basis Earthquake, t' l l Proposed Amendment No. 182, Rcvision 1 D5-5 l

.I W' ~ ' t RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS tE ADMINISTRATIVE CONTROLS o D6.1 RESPONSIBILITY'

D6.1.1 The Assistant General Manager (AGM), Nuclear shall be responsible for the management of the overall facility and ensuring the safe storage of irradiated core components. and shall delegate in writing the succession of his-responsibilities during absences, w

H D6.1.2 The Shift Supervisor-or a designated individual who is a certified fuel handler shall be responsible for the control room command function. D6.2 - ORGANIZATION D6.2.1 Offsite and'onsite organizations shall De established for corporate management and facility operation during the LONG TERM DEFUELED HODE. The organizations shall include the positions responsible for activities affecting the safe storage of irradiated core components and the overall safety of the facility. a. Lines of authority, responsibility, and communication shall be established .and defined for the highest management levels through intermediate levels, including all organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional de:;criptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. The organization and responsibilities shall be documented in the Updated Safety Analysis Report (USAR). b. The AGM, Nuclear shall have responsibility for the overall safe operation .of the facility and shall take measures necessary to ensure acceptable performance of the staff in maintaining and providing technical support to the facility to ensure NUCLEAR SAFETY. c. The individuals who train the Certified Fuel Handlers and those who carry out health physics-functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence. q

d. -The quality assurance organization shall report to the AGM, Nuclear to ensure independence from scheduler pressures.

L l D6.2.2 The facility organization shall be as shown in the USAR, and-a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table' 06.2-1. In addition, the individual who j g directly supervises the Shift Supervisors shall be a Certified Fuel . Handler. l b. The handling of spent fuel shall be directly supervised by a member of the plant _ management staff who is a Certified Fuel Handler and who has no other concurrent responsibility during a fuel handling operation. -Proposed . Amendment No.182, Revision 1 D6-1

F l (' RANCHO SECO UNIT 1 p LONG TERM DEFUELED TECHNICAL SPECIFICATIONS TABLE D6.2-1 MINIMUM SHIFT CREW REQUIREMENTS

• i L

i-Position Defueled Shift Supervisor ** 1 l Non-Certified Operator 1 Minimum Total Personnel 2 In the event that any member of a minimum shift crew is absent or incapacitated due to illness or injury, a qualified replacement shall be designated to report on site within 2 hours. - The Shift Supervisor shall be a Certified Fuel Handler l Proposed Amendment No. 182, Revision 1 D6-2

RANCHO SECO UNIT 1 LONG TERM DEFUELED i TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.3 FACILITY STAFF OUALIFICATIONS Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. D6.4 TRAINING Retraining and replacement training programs for the Certified fuel Handlers shall be maintained under the direction of the AGM, Nuclear and shall be conducted in accordance with the approved training progra:ns. J I D6.5 REVIEW AND AUD.II D6.5.1 Nuclear Safety Review and Audit Committee (NSRAC) D6.5.1.1 The NSRAC shall provide independent review and audit on matters related to NUCLEAR SAFETY, the safe storage of irradiated core components, and l ensuring acceptable performance of the staff in maintaining and providing technical support to the facility, and advise the AGM, Nuclear on these l matters. D6.5.1.2 The NSRAC shall be comprised of a Chairman and a minimum of six l members. The NSRAC membership shall be made up of personnel filling positions l l .in the Nuclear Organization who meet or exceed the minimum qualifications of ANSI N18.1-1971, Section 4.2 or 4.4, and at least 2 personnel are from outside l l the line nuclear organization end meet or exceed the minimum qualifications of ANSI /ANS 3.1-1981, Section 4.7.2. D6.5.1.3 The Chairman of the NSRAC shall be the Deputy AGM, Nuclear. A i current list of members of the NSRAC shall be developed and approved by the AGM, Nuclear. In the absence of the NSRAC Chairman, an alternate Chairman may be designated by the NSRAC Chairman or the AGM, Nuclear. The alternate Chairman shall be selected from the current list of NSRAC members. 3 D6.5.1.4 A list of alternate NSRAC members shall be developed and approved by the AGM, Nuclear. D6.5.1.5 The NSRAC shall meet at least once every other calendar month, as convened by the NSRAC Chairman, or more frequently as directed by the AGM, Nuclear. D6. 5.1. 6 A quorum of the NSRAC shall consist of a majority of the current i NSRAC members, including the Chairman or alternate Chairman. No more than two alternates shall participate in NSRAC activities at any one time for the 1 purpose of establishing a quorum. I Proposed j Amendment No. 182, Revision 1 D6-3 j

L RANCHO SECO UNIT 1 r LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADtillilSTRATIVE CONTROLS D6.5.1.7 The NSRAC shall be responsible for review of: a. The safety evaluations performed on proposed changes, tests, or experiments under the provisions of 10 CFR 50.59 that affect NUCLEAR SAFETY. ti. Proposed changes, tests, or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59. c. Proposed changes to the Technical Specifications or the Facility License. - NOTE - Changes, tests and experiments, which are determined by a Qualified Reviewer and a second level review to noi involve an unreviewed Safety Question, a change to the Facility License or Technical Specifications, or a change to a licensing basis document, are not j required to be reviewed by the NSRAC, except when otherwise stated in a Specification, d. The safety evaluations for (1) all procedures, plans, manuals, and programs required by Specification D6.8 and changes thereto, and (2) any other procedures, plans, manuals, and programs or changes thereto which l are determined by the AGM, Nuclear to affect NUCLEAR SAFETY. e. Investigations of violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, internal l procedures or internal instructions having NUCLEAR SAFETY significance, f. Significant deviations from normal and expected performance of plant l equipment that affect NUCLEAR SAFETY or represent a potential safety

hazard, g.

LICENSEE EVENT REPORTS as defined by 10 CFR 50.73, 73.71, and 26.73 and NUREG-1022 to determine adequacy of corrective action and to detect a j degrading trend. h. Any-special-investigation or facility activity brought to the attention of the NSRAC by the District's executive management which may be indicative i of conditions adverse to NUCLEAR SAFETY. 1. Major changes to Radioactive Haste Treatment Systems (Liquid, Gaseous and Solid). 4 j. Any accidental, unplanned, or uncontrolled release of radioactive material I to the environs, including the reports covering the evaluation, recommendations, and corrective actions taken to prevent recurrence. The NSRAC shall forward these reports to the AGM, Nuclear. l Proposed Amendment No.182, Revision 1 D6-4

L RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.5.1.8 The NSRAC shall: a. Report to and advise the AGM, Nuclear of activities in those areas of responsibility specified in Specification D6.5.1.7. b. Render determinations in writing with regard to whether or not each item i considered under Specifications D6.5.1.7a through 7e, and 71 above constitutes an unreviewed safety question, c. Recommend to the AGM, Nuclear other areas of facility activities where ) additional auditing is prudent. j d. Advise the AGM, Nuclear of the need for independent auditing of facility l activities. e. Provide oversight review of: 1) The Quality Assurance Program, including content, implementation, and reports. 2) The radiological waste and effluent control programs. 3) The Licensee Event Report program. 06.5.1.9 The NSRAC may form subcommittees to screen reviews required by l Specification D6.5.1.7. D6.5.1.10 Minutes of each NSRAC meeting and separate documentation of reviews performed per Specifications D6.5.1.7c, e, f, and g shall be prepared, approved, and forwarded to the AGM, Nuclear within 14 days following each i

meeting, D6.5.2 -NOT USED-D6.5.3 Technical Review and Control l

c l Activities which affect NUCLEAR SAFETY shall be conducted as follows: i a. Procedures, plans, manuals, and programs required by Specification D6.8 and other procedures, plans, manuals, and programs which affect NUCLEAR SAFETY, and changes thereto, shall be prepared, reviewed, and approved. Each such procedure, plan, manual, and program or change thereto shall be reviewed by an individual (s) other than the preparer, but who may be from the same organization as the preparer of the procedure, plan, manual, and program or change thereto. Programs, plans, manuals, and procedures other than plant administrative procedures will be approved as delineated by specific technical specifications, otherwise, as delineated in writing by l the AGM, Nuclear, but not lower than the manager level (the manager level includes Area Heads and higher as defined in plant administrative i l procedures). Such procedures, plans, manuals, and programs shall be reviewed periodically in accordance with administrative procedures. l l l l Proposed l Amendment No. 182, Revision 1 o D6-5 i

L RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTRQLS D6.5.3 Technical Review and Control (Continued) The AGM, Nuclear will approve plant administrative procedures, Security Plan Implementing Procedures, and Emergency Plan Implementing Procedures. Approval of temporaTy procedure, plan, manual, or program changes which clearly do not change the intent of the approved procedure, plan, manual, or program can be made by two members of the plant management staff, at least one of whom is a Certified fuel Handler or NSRAC member, as a l minimum. The change shall be documented, reviewed, and approved by the designated approval authority for that procedure, plan, manual, or program within 14 days of implementation, b. Proposed changes or modifications to plant systems or equipment that affect NUCLEAR SAFETY shall be reviewed by an individual (s) other than the individual (s) who designed the modification, but who may be from the same organization as the individual (s) who designed the modifications. Such modifications shall be approved by the AGM, Nuclear or his designee as delineated in writing, but not lower than the manager level. c. Proposed tests and experiments which affect NUCLEAR SAFETY and are not addressed in the USAR shall be reviewed by an individual (s) other than the individual (s) who prepared the proposed test or experiment. Such tests or experiments shall be approved by the AGM, Nuclear or his designee as delineated in writing, but not lower than the manager level, d. Individuals responsible for reviews performed in accordance with Specification 06.5.3a, b, and c shall meet or exceed the qualification requirements of Section 4.4 of ANSI N18.1-1971. Each such review shall include a determination of whether or not additional, cross-disciplinary review is necessary. A list of qualified reviewers for the independent reviews described in 06.5.3a, b, and c above shall be established by the AGM, Nuclear. The personnel performing the cross-disciplinary review need not be qualified reviewers, but a qualified reviewer shall review each determination, e. Events reportable pursuant to Technical Specification D6.9.4 and violations of Technical Specifications shall be investigated and a report prepared which evaluates the event and which provides recommendations to l prevent recurrence. Such reports shall be reviewed by the NSRAC and l forwarded to the AGM, Nuclear. D6.5.4 Audits Audits of facility activities shall be performed under the cognizance of the l Manager, Nuclear Quality / Environmental Monitoring and Industrial Safety. These audits shall encompass: a. The conformance of facility activities to provisions contained within the Technical Specifications and applicable license conditions at least once per year. l Proposed Amendment No. 182, Revision 1 D6-6 l

i RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6_.5.4 Audits (ContinuedF b. The training and qualifications of the District's facility technical staff !l at least once per year. c. The result of actions taken to correct deficiencies occurring in facility equipment, structures, systems or methods of conducting shutdown activities that affect NUCLEAR SAFETY at least once per year. d. The performance of activities required by the Quality Assurance Program at least once per 2 years, j e. The Facility Emergency Plan and implementing procedures at least once per 2 yeart l f. The Facility Security Plan and implementing procedures at least once per year. g. Any other area of facility activity considered appropriate by the AGM,

Nuclear, h.

Compliance with fire protection requirements and implementing procedures at.least once per 2 years.

i. An independent fire protection and loss prevention program inspection and audit shall be performed annually using either qualified offsite licensee personnel or an outside fire protection firm. A qualified outside fire protection firm shall be used at least once every 3 years.

'j. The REMP results at least once per year, k. The ODCM and REMP Manual and implementing procedures thereof at least once per 2 years. 1. The PCP and implementing procedures for processing and packaging of radioactive wastes from liquid systems at least once per 2 years. m. The performance of activities required by the Quality Assurance Program for Effluent Control and Environmental Monitoring at least once per year. Audit reports of reviews encompassed by Specification D6.5.4 shall be forwarded to the AGM Nuclear and to the management positions responsible for the areas reviewed within 30 days after completion. D6.6 -NOT USED-D6.7 -NOT USED- _ Proposed Amendment No. 182, Revision 1 06-7 i

I RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMIN':STRATIVE CONTROLS D6.8. PROCEDURES. PLANS. MANUALS. AND PROGRAMS D6.8.1 Hritten procedures / plans / manuals / programs shall be established, implemented, and maintained covering the activities referenced below: a. The procedures recommended in Appendix "A" of Safety Guide 33, November 1972 applicable in the LONG TERM DEFUELED MODE. b. The safe storage of irradiated core components c. Surveillance and test activities on equipment required for long term safe storage of irradiated core components d. Security Plan implementation e. Emergency Plan implementation f. Fire Protection Program Plan implementation g. PCP implementation h. ODCM implementation i. REMP MANUAL implementation

j. Quality Assurance for Effluent Control and Environmental Monitoring using the guidance of Regulatory Guide 4.15, Revision 1. February 1979 k.

Certified Fuel Handler Training Programs implementation 1. Quality Assurance Program Implementation D6.8.2 Each procedure / plan / manual / program of Specification D6.8.1 above and changes thereto shall be reviewed and approved as set forth in Specification 06.5. Additionally, changes to the ODCM and REMP MANUAL shall be processed in accordance with Specification D6.14, and changes to the PCP shall be processed 3 in accordance with Specification D6.13. D6.8.3 The following programs shall be established, implemented, and i maintained: a. Radioactive Effluent Controls Program l A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBER (S) 0F THE PUBLIC from radioactive effluents as low as reasonably achieveable. The program (1) shall be contained in the ODCM, (2) shall be implemented by Administrative, Chemistry, and Operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, i 4 2) Limitations on the concentrations of radioactive material released in j liquid effluents to unrestricted areas conforming to 10 CFR Part 20, ( Appendix B. Table II, Column 2, i 1 Proposed Amendment No. 182, Revision 1 D6-8

I RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.8 PROCEDURES. PLANS. MANUALS. AND PROGRAMS (Continued) 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM, l 4) Limitations on the annual and quarterly doses or dose commitment to MEMBER (S) 0F THE PUBLIC from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50, 5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in l the ODCM at least every 31 days, 6) Limitations on the OPERABILITY and use of the gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 7) Limitations on the OPERABILITY and use of the liquid effluent treatment system to ensure that the appropriate portions of this system is used to reduce releases of radioactivity when the projected doses in a-31-day period would exceed 8 1/3 percent (1/12) of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR 50. 8) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR 20, Appendix B Table II, Column 1, 9) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix ; to 10 CFR Part 50, 10) Limitations on the annual and quarterly doses to a MEMBER (S) 0F THE PUBLIC from Tritium and all radionuclides in particulate form with l half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 11) Limitations on the annual dose or dose commitment to MEMBER (S) 0F THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190. Proposed Amendment No. 182, Revision 1 D6-9

I RANCHO SECO UNIT 1 4 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.8 PROCEDURES. PLANS. MANUALS. AND PROGRAMS (Continued) b. Radioloaical Environmental Monitorina Proaram (REMP) A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the REMP MANUAL, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following: 1) Monitoring. sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the REMP manual, 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and 3) Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring, i D6.8.4 Each program of Specification D6.8.3 above and changes thereto shall be reviewed and approved as set forth in Specification 06.5. Additionally, changes to the ODCM and REMP MANUAL shall be processed in accordance with Specification 06.14, and changes to the PCP shall be processed in accordance with Specification D6.13. l D6.9 REPORTING REOUIREMENTS In addition to the applicable reporting requirements of Title 10 to the Code of Federal Regulations the following reports shall be submitted to the Regional Administrator of the Region V Office unless otherwise noted, p D6.9.1 Annual Radiological Recorts l l Annual reports covering the activities of the unit, as described below, for the previous calendar year shall be submitted as follows: 06.9.1.1 Annual Occuoational Radiation Exoosure Report l The Annual Occupational Radiation Exposure Report for the previous calendar year shall be submitted to the Commission within the first calendar quarter of each calendar year in compliance with 10 CFR 20.407. Proposed l kaendment No.182, Revision 1 l 06-10 l

RANCHO SECO UNIT 1 l LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.9.1.2 Annual Exoosure Repp.tt The Annual Exposure Report for the previous calendar year shall be submitted to the Commission within the first calendar quarter of each calendar year in accordance with the guidance contained in Regulatory Guide 1.16. D6.9.1.3 Annual Radioloaical Environmental Ooeratina Reoort The Annual Radiological Environmental Operating Report covering the activities of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The content of this report shall be consistent with the objectives outlined in (1) the REMP MANUAL and (2) Sections IV.B.2 IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. D6.9.2 Semiannual Radioactive Effluent Release Reoort The Semiannual Radioactive Effluent Release Report covering the activities of the unit during the previous 6 months shall be submitted within 60 days after January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The content of this report shall be consistent with the objectives outlined in (1) the ODCH and PCP, (2) in conformance with 10 CFR 50.36a, and (3) consistent with Section IV.B.1 of Appendix I to 10 CFR Part 50. D6.9.3 Annual Reoort A routine report consisting of shutdown statistics, a narrative summary of f shutdown experience and major maintenance of equipment required for long term safe storage of irradiated core components, and tabulations of facility changes, tests, or experiments required pursuant to 10 CFR 50.59(b) shall be submitted on an annual basis to the U. S. Nuclear Regulatory Commission, Document Control Desk, Washington, D. C. 20555, with a copy to the Regional Office, postmarked no later than 30 days following the twelve month period covered by the report. Proposed Amendment No. 182, Revision-1 ) 06-11

RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.9.4 License.e Event Recort D6.9.4.1 The types of events listed in 10 CFR 50.73, 73.71, and 26.73 shall be the subject of Licensee Event Reports, submitted to the U.S. Nuclear Regulatory Commission (NRC), Document Control Desk Washington, D.C. 20555, within the time requirements of 10 CFR 50.73. An additional copy shall also be submitted to the Regional Administrator of the Region V Office. The written report shall include a completed copy of a Licensee Event Report form, pursuant to 10 CFR 50.73 and the guidance of NUREG-1022,.nd a description of corrective actions and measures that are designated to prevent recurrence. Supplemental reports may be required to fully describe final resolution of the occurrence. For corrected or supplemental reports, another Licensee Event Report shall be completed, and reference shall be made to the original report date, pursuant to the requirements of 10 CFR 50.73, 73.71, and 26.73. D6.9.4.2 The following actions shall be taken for events which are reportabit as Licensee Event Reports: a. The NRC shall be notified and a report submitted pursuant to the requirements of 10 CFR 50.73, 73.71, and 26.73, and b. Each Licensee Event Report shall be reviewed by the NSRAC. Also, each Licensee Event Report shall be reviewed and approved by the AGM, Nuclear, or designee. D6.9.5 -NOT USED-D6.9.6 Environmental Reporti a. When a change to the plant design or to plant activities is planned which would have a significant adverse effect on the environment or which involves an environmental matter or question not previously reviewed and evaluated by the NRC, a report on the change will be made to the NRC prior to implementation. The report will include a description and evaluation of the change, including a supporting benefit-cost analysis. b. Changes or additions to permits and certificates required by Federal, State, local, and regional authorities for the protection of the environment will be reported. When the required changes are submitted to the concerned agency for approval, they will also be submitted to the NRC for information. The submittal will include an evaluation of the environmental impact of the change, r / Proposed Amendment No.182, Revision 1 D6-12

a RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.10 RECORD RETENTION D6.10.1 The following records shall be retained for at least 5 years: a. Records and logs of facility operation covering time interval at each power level. b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to NUCLEAR SAFETY. c. Licensee Event Reports, d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications. e. Records of reactor tests and experiments. l f. Records of changes made to Operating Procedures, g. Records of radioactive shipments, h. Records of sealed-source leak tests and results. i. Records of annual physical inventory of all sealed-source material of record. j. Records and logs of facility activities in the Long Term Defueled Mode. l 06.10.2 The following records shall be retained for the duration of the Facility License: a. Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis

Report, b.

Records of new and irradiated fuel ir,ventory, fuel transfers, and assembly burnup histories, c. Records of facility radiation and contamination surveys, d. Records of radiation exposure for all individuals entering radiation control areas. e. Records of gaseous and liquid radioactive material released to the

environs, f.

Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles. Proposed Amendment No. 182, Revision 1 o 06-13

RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.10 RECORD RETENTION (Continued) g. Records of training and qualification for current members of the plant operating staff. h. Recc ds of in-service inspections performed previously pursuant to the Appendix A Technical Specifications. 1. Records of quality assurance activities required by the 0A Manual. _j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59. k. Records of meetings of the PRC and MSRC. 1. Records for Environmental Qualification of safety-related electrical equipment previously performed pursuant to Appendix A Technical Specification 6.14, and records of changes made to the Environmental Qualification of this equipment in the LTDH. ] m. Records for the Radiological Environmental Monitoring Program, n. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records, o. Records of reviews performed for changes made to the ODCH, REMP HANUAL, and the PCP. p. Records of plant closure activities performed. q. Records of meetings of the NSRAC. D6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 19 and 10 CFR 20, and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. L Proposed Amendment No. 182, Revision 1 06-14

I RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS D6.12 HIGH RAQIATION ARIA D6.12.1 In lieu of the " control device" or " alarm signal" required by 10 CFR 20.203(c)(2), a. Each High Radiation Area in which the intensity of radiation is equal to + or greater than 100 mrem /hr but less than 1,000 mrem /hr shall be barricaded and conspicuously posted as a High Radiation Area, and entrance thereto shall be controlled by issuance of a Radiation Hork Permit. Any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area, b. Each High Radiation Area in which the intensity of radiation is equal to or greater than 1,000 mrem /hr shall be subject to the provisions of Specification D6.12.la above, and, in addition, locked doors shall be provided to prevent unauthorized entry into such area. The keys shall be maintained under the administrative control of the on duty Shift Supervisor. In lieu of locked doors, certain areas within the Reactor Building may employ conspicuous visible or audible signals such that an individual is made aware of the presence of the High Radiation Area (>1,000 mrem /hr). D6.13 PROCESS CONTROL PROGRAM (PCP) D6.13.1 The required content of the PCP is defined in Specification D1.6. D6.13.2 Changes to the PCP: a. Shall be documented and records of reviews performed shall be retained as required by Specification D6.10.20. This documentation shall contain: 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s), and 2) A determination that the change will maintain the overall conformance of the processed waste product to existing requirements of Federal, State, or other applicable regulations, b. Shall become effective after review and acceptance by the NSRAC and l approval by the AGH, Nuclear. 1 l Proposed Amendment No. 182, Revision 1 D6-1S

i RANCHO SECO UNIT 1 LONG TERM DEFUELED TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS i R6 14 0FFS11E DOSE CALCULATION AND RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MAN MLS D6.14.1 The required content of the ODCM is defined in Specification 01.7. D6.14.2 The required content of the REMP MANUAL is defined in D1.9. D6.14.3 Changes to the ODCM or REMP MANUAL: a. Shall be documented and records of reviews performed shall be retained as required by Specification D6.10.20. This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent dose, or setpoint calculations, b.

Shall become effective after review and acceptance by the NSRAC and l i approval by the AGM, Nuclear, c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCH and/or REMP MANUAL as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM and/or REMP HANUAL was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented. l l l Proposed Amendment No. 182, Revision 1 06-16 1

I r 5' N L ATTACHMENT II SATETY ANALYSIS REPORT j

I License Change Safety Analysis Log No. 1091, Revision 1 Proposed Amendment No. 182, Revision 1 Page I Cf 38 L Reicription_oLChange Proposed Amendment No. 162, Revision 1 (PA-182, Rev. 1) creates a new L appendix, Appendix C, to Facility Operating License No. DPR-54. Appendix C, or Long Term Defueled Technical Specifications (LTDTS), will be followed in lieu of the existing Appendix A and B Technical Specifications for the current extended outage that began in June 1989, or until NRC approval of a Decommissioning Plan and the facility license is amended commensurate with a permanently shutdown facility. The LTDTS define a new mode, titled Long Term Defueled Mode (LTDM), which applies to Rancho Seco's current defueled condition. Movement of fuel into the Reactor Building would take the plant out of the LTDM and the LTDTS (Appendix C), and cause the plant to revert back to the Appendix A and B Technical Specification requirements. The LTDTS provide Limiting Condition and Surveillance requirements for the systems and components required to be OPERABLE to ensure long term safe storage of irradiated core components in the spent fuel pool. The LTDTS also provide Administrative Control requirements for the overall safe operation of the facility. Implementation of the LTDTS will ensure the continued protection of the health and safety of the pub".ic during the LTDM. Reason for change The District developed the proposed LTDTS based on (1) Rancho Seco's current defueled condition, and-(2) the District's acknowledged intent to nat operate Rancho Seco in the future. The fuel is removed from the reactor vessel and is f now stored in the spent fuel pool (SFP). Plant systems are being layed up in either a wet or dry condition, as appropriate, to limit degradation during the l current extended outage. The-LTDTS provide (1) the level of plant protection needed to deal with the credible accidents postulated in the defueled condition, (2) long term safe storage of the nuclear fuel, and (3) the necessary level of protection of the public's health and safety. The LTDTS provide a clear and concise set of Specifications applicable in an extended outage and defueled condition. The LTDTS provide relief from those Appendix A & B Technical Specification requirements which do not apply in the LTDM. The LTDTS are submitted as an addition to the current Technical Specifications and are intended to be Appendix C to facility Operating License No. DPR-54. Upon NRC approval, the LTDTS become applicable in lieu of the Appendix A and B Technical Specifications while the plant is in the LTDM or until the facility license is amended otherwise. l-l 1 i j

[ License Change Safety Analysis Log No.1091, Revision 1 Proposed Amendment No. 182, Revision 1 Page 2 cf 38 [ n luglion and Basis For Safety FindlDgi The primary difference between the proposed LTDTS and the current Rancho Seco Technical Specifications is the range of postulated accidents that these two sets of Specifications are required to address, and the range of potential radiological consequences resulting from abnormal plant conditions in the various applicable modes. Based on a review of these conditions and credible . accidents, the LTDTS were developed to provide the necessary level of protection. The Updated Safety Analysis Report (USAR), Chapter 14, contains two accidents or conditions that are considered credible in the LTDM: 1. Fuel Handling Accident 2. Complete Loss of All Unit a-c Power (LOOP) These two accidents or conditions apply in the defueled condition and are i considered in the analysis below. Fuel Handlina Accident The spent fuel at Rancho Seco has undergone substantial decay (294 days as of March 28, 1990), thus the source term for a credible accident in the LTDM is significantly less than that assumed for an operating plant. Essentially no radioactive iodine is present in the plant that could have a significant dose impact on members of the public, thus the calculated thyroid exposure levels resulting from credible accidents in the LTDM are negligible. The primary l-radioisotope of concern from an offsite exposure (dose) standpoint following i the dropped fuel assembly accident, the maximum credible accident in the LTDM, is Krypton-85. Results of the analysis for the dropped fuel assembly accident in the current defueled condition show that the 2-hour integrated total body dose attributed to the maximum exposed individual is 0.013 rem. This calculated dose is an extremely small fraction of the 10 CFR 100 accident dose limit (25 rem), and is significantly less than the annual dose limit of l 10 CFR 20 (0.5 rem). The District calculation is supported by the safety evaluation that accompanied current Technical Specification 3.13.2 (Amendment 39) which requires only a 30 day spent fuel decay time prior to L fuel movement when the Auxiliary and Spent Fuel Building filter systems are not OPERABLE. The District expects only limited fuel handling activities l while the plant is in the LTDM. 1

i I License Change Safety Analysis Log No.1091, Revision 1 p Proposed Amendment ko.182, Revision 1 Page 3 Cf 38 i LDQE During normal plar.t operation; and post accident conditions it is imperative that electrical power be available to support equipment needed to operate the plant and mitigate the consequences of an accident. In the LTDM, a LOOP would result in the loss of the spent fuel pool cooling (SFC) system.

However, there is adequate time available (see the SFP decay heat load evaluation below) to take corrective action without a safety consequence in the event of a LOOP.

Rancho Seco has six offsite power transmission lines, and has the capability to receive power directly from either the District's or PG&E's hydroelectric units in less than eight hours. An evaluation of the offsite electrical grid for Rancho Seco, performed pursuant to 10 CFR 50.63, Station Blackout, verified the stability of the Western grid. The probability of a LOOP at Rancho Seco, as evaluated in accordance with the guidelines of Regulatory Guide 1.155, is less than once per 20 years. Therefore, the emergency diesel generators are not required to ensure power availability to support SFC equipment in the event of a LOOP. An alternate power supply can be made available well within the minimum time required to take corrective action to restore SFC. A LOOP will not result in a release of any significant amount of SFP inventory or radioactivity. SFP Decav Heat Load The controls required to protect the spent fuel in the LTDM are predicated primarily on the level of decay heat in the SFP. The decay heat load for the SFP in the defueled condition was calculated using the methodology described in ANSI /ANS 5.1-1979 and BTP ASB 9-2. The decay heat load in the SFP, as a function of time, is shown in Table 1 (see page 34 of this Safety Analysis Report). Table 1 also provides, as a function of calendar date, the time required for the SFP to reach 2;2'F from an initial temperature of 120'F, and the time required to boil 6.75 feet of water from the SFP following a loss of SFC. The 6.75 feet was chosen since the SFP water level is normally maintained above 37 feet (the low-low level alarm) and, as evaluated previously, must be maintained at 30 feet 3 inches to limit the dose rate at the SFP surface to 2.5mr/hr when 1080 spent fuel assemblies that have decayed 3 days following irradiation at 100% power is stored in the SFP. This SFP surface dose rate calculation is very conservative since the reactor was not operated at 100% power as assumed in the dose rate calculation, the number of spent fuel assemblies stored in the SFP is much less than 1080, and the actual decay time of the 493 spent fuel assemblies in the SFP is significantly higher than the 3 days assumed. No safety implication exists for a significantly lower SFP level as long as personnel exposure is monitored and maintained as low as is reasonably achievable. LTDTS Specification D3/4.1 provides limits for the SFP surface dose rate. The results. presented in Table 1 are conservative since they are based on the following conservative assumptions: 1. The power history for nach reactor run is calculated to have been accumulated as the effective full power days (EFPD) of the run at 100% power at the end of the run (i.e., the 245 EFPD run ending on June 7,1989 is assumed to have occurred at 100% power in the last 245 days before June 7, 1989 when in fact it was spread out over a 48 month period).

License Change Safety Analysis Log No.1091, Revision 1 Proposed Amendment No. 182, Revision 1 Page 4 cf 38 f 2. The decay heat load calculations include a 101, conservatism factor. 3. There is no heat loss from the SFP due to evaporation or ambient losses, except after the SFP reaches the boiling point. As seen in Table 1, upon a loss of SFC the minimum time required to reach boiling in the SFP is 74.4 hours as of January 3,1990, and increases to 96.12 hours on June 7, 1990. Thus, a minimum of 3 days, and in the near future 4 days, is available to restore SFC prior to the occurrence of boiling following a loss of SFC. In addition, if the extra time available through boil-off of 6.7S feet of the 37 foot normal minimum water depth is taken into account, a minimum of 9.18 days as of January 3,1990 and 11.86 days as of June 7,1990 is available for corrective actions to be implemented to restore SFC. If e boil-off of SFP water were to occur due to a loss of SFC, a simple addition of water to the SFP would extend the time to implement corrective actions to restore SFC, if necessary. No degradation effects on the fuel and cladding is associated with SFP boiling since the fuel is designed to operate at significantly higher coolant temperatures than 212'F. The effect of thermal stresses on the structural support of the SFP have been analyzed and found acceptable up to a SFP temperature of 212*F. Therefore, a 212'F SFP temperature does not impose a safety hazard. Criticality Control Criticality control in the SFP is achieved through the use of high density fuel storage racks that contain a neutron absorbing material (Boraflex) within the rack plates. These racks are designed to hold 1,080 fuel assemblies at 4,0 weight percent enrichment with unborated water in the SFP while still maintaining Kett less than 0.95. The current fuel inventory in the SFP is 493 assemblies with a maximum enrichment prior to burnup of 3.21 weight percent. District calculations show that the estimated maximum synthesized enrichment (remaining U-235 plus fissile Pu) for any assembly in the SFP is 2.573 weight percent, thus providing an even greater shutdown margin. Also, control assemblies are stored in several spent fuel assemblies. This provides additional neutron absorption material; therefore, no boron is required in the .SFP to maintain adequate shutdown margin. The SFP storage rack neutron absorption capability is monitored through a coupon sampling program. Boraflex material samples are periodically removed l from the SFP for testing to verify rack performance. To date, the removed l coupons have provided verification of the Boraflex integrity, absorption capability, and spent fuel storage rack performance as designed. In addition, the LTDTS contain new requirements to monitor the SFP water for chloride and fluoride levels to assure rack and fuel structural integrity is maintained over time. The combination of these two programs provides assurance that the spent fuel storage racks will remain structurally sound and perform their i intended. criticality control function. 1 l

N ? 4h License Change Safety Analysis Log No. 1091, Revision 1 Proposed Amendment No. 162. Revision 1 Page 5 cf 38 An additional' event considered for the LTDM is the drop of a spent fuel cask. This event is addressed in USAR sections 9.8.2.4 and 9.8.2.5. Because of the many mechanical and electrical interlocks and administrative controls associated with the Turbine Building Gantry Crane, which are designed to prevent movement of a cask over spent fuel assemblies, the drep of a cask onto spent fuel assemblies is not considered a credible event. In addition, plant configuration and crane interlocks prevent a spent fuel cask from being lifted to a height more than 26 feet in air or 33 feet 6 inches in water during the cask transport process. Casks are required to be rated for a 30 foot drop in air pursuant to 10 CFR 71. This height in air is equivalent to more than a 40 foot drop in water. Therefore, cask integrity is assured in the unlikely event a cask is dropped. i In addition to the above evaluation of the accidents and conditions considered credible in the LTDM, several Generic Letters, as listed below, are considered in the preparation of the LTDTS as follows: a. Generic Letter 89-01, IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS IN THE ADMINISTRATIVE CONTROLS SECTION OF THE TECHKICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF RETS TO THE OFFSITE DOSE CALCULATION MANUAL OR THE PROCESS CONTROL PROGRAM b. Generic Letter 89-14 LINE-ITEM IMPROVEMENTS IN TECHNICAL SPECIFICATIONS - REMOVAL OF THE 3.25 LIMIT ON EXTENDING SURVEILLANCE INTERVALS c. The previously docketed submittals made ptrsuant to Generic Letter 88-12. REMOVAL OF FIRE PROTECTION REQUIREllENTS FROM TECHNICAL SPECIFICATIONS, are acknowledged in this submittal by omission of the appropriate Technical Specifications. Proposed Amendment No.180 provides the evaluation and justification for the removal of fire protection related Technical Specifications pursuant to Generic Letter 88-12. The following is a line by line discussion of the current Technical Specifications and their disposition in relationship to the LTDTS. Technical Specification Number Eiscussion 1.1 The term Rated Power is not applicable in the LTDM and is not included in the LTDTS. 1.2 Definitions 1.2.1 through 1.2.12 and 1.2.14, and Table 1.2-1 are related to plant operations or non defueled activities, are not l applicable in the defueled condition, and are not included in the LTDTS. Specification 1.2.13 (Action) is included as LTDTS D1.1. A definition for the LTDM is included as LTDTS D1.2. i

~ License Change Safety Analysis Log No. 1091, Revision 1 Proposed Amendment No. 182, Revision 1 Page 6 of 38 Technical. Specification Number Discussion 1.3 The definition for the term OPERABLE is modified to remove the requirement for an operable emergency power source while the plant is in the LTDM. The current Technical Specifications do not require the emergency power supplies to be OPERABLE when the plant is in a mode below Heatup-Cooldown. Also, the requirement for an OPERABLE redundant system is not applicable in the LTDM. A revised definition of OPERABLE is included in the LTDTS as Specification D1.3. 1.4 These definitions apply to systems such as SFAS and RPS that are not in service in the LTDM. These terms are not used in the LTDTS and are not included in the Definition section. 1.5 The definitions from this Specification that are used in the j LTDTS, i.e. Channel Test, Instrument Channel Check, Instrument Channel Calibration, and Functional Test, are included as Definitions DI.4.1, DI.4.2, Dl.4.3 and D1.4.4, respectively. The remaining terms are not used in the LTDTS and are not included in the Definition section. 1.6 This definition section describes core reactivity conditions that are not applicable with the reactor defueled and are therefore not included in the LTDTS. l7 This Specification defines Containment Integrity which is not applicable in the LTDM. Therefore, this definition is not included in the LTDTS. 1.8 The description for Licensee Event Report (LER) is included in the LTDTS as specification D6.9.4. Therefore, a definition for LER placed in the LTDTS Definition section is not necessary and is not included. 1.9 The surveillance intervals applicable for the LTDTS are included as Speci fication D1.5 and Table D1.5-1. The requirement to not exceed 3,25 surveillance time intervals for any three consecutive { intervals is removed and rewritten in accordance with Generic Letter 89-14. 1.10 The term Safety is changed to Nuclear Safety and is included in the LTDTS as Specification 01.10. 1.11 This Specification defines Fire Protection terms that are removed in Proposed License Amendment No. 180 per Generic Letter 88-12. Therefore, these terms are not included in the LTDTS.

License Change Safety Analysis Log No. 1091, Revision 1 Proposed Amendment No. 182. Revision 1 Page 7 cf 38 Technical Specification Number Discussion 1.12 This definition, Staggered Test Basis, is not used in the LTDTS and thus is not included. 1.13 The definition for the Process Control Program (PCP) is included in the LTDTS as Definition 01.6 and has been modified per the guidelines of Generic Letter 89-01. 1.14 The term Solidification has been replaced by a more generic term. Processing, and relocated to the PCP per the guidelines of Generic Letter 89-01. The term Processing allows the use of new techniques that meet NRC and State requirements for disposal of wastes. 1.15 The definition for the Offsite Dose Calculation Manual (0DCM) has been modified to meet the guidelines of Generic Letter 89-01 and included in the LTDTS as Specification D1.7. 1.16 The term Restricted Area is not used within the body of the LTDTS and is therefore not included in the LTDTS Definition section. 1.17 The term Site Boundary is used in conjunction with other terms in the LTDTS and does not have a specific definition. Therefore, this term is not included in the LTDTS. 1.18 The definition for Dose Equivalent I-131 is relocated to the ODCM per Generic Letter 89-01. 1.19 The definition Member (s) Of The Public is included in the LTDTS as Specification 01.8. 1.20 The definition for Dewatering has been replaced by a more generic term, Processing, and relocated to the PCP per Generic Letter 89-01. The term Processing allows the use of new techniques that meet NRC and State requirements for disposal of wastes. 1.21 This definition, Maximum Exposed (Hypothetical) Individual, is relocated to the ODCM per Generic Letter 89-01. l 1.22 The definition for the Radiological Environmental Monitoring Program (REMP) Manual has been modified per Generic Letter 89-01 and is included in the LTDTS as Specification D1.9. -1.23 This definition, Liquid Effluent Radwaste Treatment System, is relocated to the ODCH per Generic Letter 89-01. i 1.24 The term Ventilation Exhaust Treatment Systems is relocated to the ODCM per Generic Letter 89-01. l

f License Change Safety Analysis Log No. 1091, Revision 1 Proposed Amendment No. 182 Revision 1 Page 8 cf 38 Technical Specification Number Discussion 1.25 This definition, Purge - Purging, is relocated to the ODCM per Generic Letter 89-01. 1.26 This definition, Venting, is relocated to the ODCM per Generic Letter 89-01. l 1.27 This definition, E-BAR, is for radionuclide concentrations in the RCS based on beta and gamma activity, is not applicable in the LTDM, and is not included in the LTDTS. 2 Section 2 of the Appendix A Technical Specifications provides limits for reactor and RCS operation. This section is designed to maintain the integrity of the fuel cladding during reactor operation and prevent fission product release to the RCS. These Specifications are not applicable in the LTDM. Therefore, this section is not included in the LTDTS. 3.0.1 This Specification has been modified to reflect the presence of only one mode in the LTDTS. 3.0.2 The phrase " Limiting Condition for Operation (LCO)" is changed to " Limiting Condition (LC)" since the plant is not permitted to operate in the LTDM in accordance with the LTDTS. 3.0.3 This Specification provides the time limit within which a mode change must be made when an LCO is not met. This Specification is not applicable when the LTDTS are applicable since the plant can only be in one mode. Therefore, this Specification is not included in the LTDTS. 3.0.4 This Specification provides restrictions for entry into an operational mode. This Specification is not applicable when the LTDTS are applicable since the plant can only be in one mode. Therefore, this Specification is not included in the LTDTS. 3.1 Specifications 3.1.1.1 through 3.1.9.2 are not required to be met below the Heatup-Cooldown mode and are not incluGed in the LTDTS. These Specifications are not applicable since they provide protection of the RCS and the reactor core which is not a safety concern in the LTDM. 3.2 Specification 3.2.1 provides for adequate boration control and ensures the ability to bring the reactor to cold shutdown following an accident. Since the heat and criticality source (the core) has been removed, no need to provide this type of protection exists. The requirements of Specification 3.2.1 are not included in the LTDTS. l 1 L

[ I License Change Safety Analysis -Log No. 1091, Revision-1 Proposed Amendment No. 182, Revision 1 Page 9 cf 38 Technical Specification 4 Number Discussion 3.2 During power plant operations Technical Specification 3.2.2 (Cont.) assures, in cold conditions, that the core will be covered with water by preserving the physical integrity of the reactor vessel, thereby protecting the health and safety of the public. This is accomplished by specifying requirements that are designed to 1 protect the vessel from an inadvertent pressurization when the plant is in a cold operating condition, i.e., when the RCS is at or below 350' F with fuel in the reactor and the RCS not open to the Reactor Building atmosphere. With.the reactor vessel defueled and RCS operation no longer necessary or desired, overpressure protection, to.the extent defined in Specification 3.2.2, is not required or needed to mitigate the consequences of a design basis accident considered credible in the LTDM. In the LTDM, 1 overpressure protection is needed only to help protect and preserve plant equipment for possible future use, not'to ensure the health and safety of the public. Because overpressure protection for the RCS is not required to ensure Nuclear Safety in the LTDM, overpressure protection requirements are not included in the LTDTS. The necessary level of overpressure protection in the LTDM is considered and addressed in the layup plan for the RCS. While the plant is in the LTDM, overpressure protection requirements will be administrative 1y controlled through plant procedures and will implement good engineering practices. 3.3 Specifications 3.3.1 through 3.3.2 address requirements for the Emergency Core Cooling, Reactor Building Emergency Cooling, and Reactor Building Spray systems and are not applicable below the Heatup-Cooldown mode. Since the reactor is defueled, there are no postulated accidents that require these systems to be operational. Therefore, the requirements of these Specifications are not included in the LTDTS. 3.4 Specifications 3.4.1 through 3.4.2 specify the components required to be Operable to maintain the steam generators in service as an RCS heat removal source. These Specifications are not applicable-below the Heatup-Cooldown mode since below this mode the RCS is not at a temperature where the steam generators are required to be in service. Therefore, these Specifications are not included in the LTDTS. -l i } i

I License Change Safety Analysis Log No. 1091, Revision 1 1 Proposed Amendment No.182, Revision 1 Page 10 cf 38 Technical Specification i Number Discus _sion j 3.5.1 Specifications 3.5.1.1 through 3 5.1.11 require instrumentation associated with the Reactor Protection System (RPS), Safety Features Actuation System (SFAS) Emergern Feedwater Initiation And Control System (EFIC), and other reactor operations safety equipment to be Operable for startup of the plant. These Specifications are not applicable below the Heatup-Cooldown mode. Because the reactor is defueled and no postulated accidents of any significant consequence are postulated to occur in the Reactor Building, these Specifications are not included in the LTDTS. 3.5.2 Specifications 3.5.2.1 and 3.5.2.2 provide limits on core power distribution and control rod Operability and are not applicable during the Cold Shutdown / Refueling Shutdown / Refueling Operations modes. Since the reactor is defueled, these Specifications are not included in the LTDTS. 3.5.3 Specification 3.5.3 provides setpoints for SFAS and is not applicable in the Cold Shutdown / Refueling Shutdown / Refueling Operations modes. Since the accidents mitigated by SFAS actuation are not credible in the defueled condition, this Specification is not included in the LTDTS. 3.5.4 Specifications 3.5.4.1 and 3.5.4.2 provide core axial imbalance and radial tilt limits for reactor operation. These Specifications are not applicable during the Cold Shutdown / Refueling Shutdown / Refueling Operations modes and are not included in the LTDTS since there is no fuel in the core. 3.5.5 Specification 3.5.5 provides a list of accident monitoring equipment that must be Operable for reactor operations. As stated in Note (1) of Technical Specification Table 3.5.5-1, this Specification is not applicable in the Cold Shutdown / Refueling Shutdown / Refueling Operations modes and is therefore not included in the LTDTS. The postulated accidents these instruments are required to monitor are not credible in the LTDM. 3.5.6 Specification 3.5.6 provides setpoints for the EFIC system. The EFIC system is not required to be Operable in the Cold Shutdown / Refueling Shutdown / Refueling Operations modes. This Specification is not included in the LTDTS since the steam generators that the EFIC system supports are not required in the LTDM. 3.5.7 Specification 3.5.7 provides a list of instrumentation required to achieve and maintain Hot Shutdown from outside the Control Room. This capability is required when the plant is in Hot Shutdown or above and is not required in the Cold Shutdown / Refueling Shutdown / Refueling Operations modes. Therefore, this Specification is not included in the LTDTS. a

License Change Safety Analysis Log No.1091, Revision 1 Proposed Amendment No. 182, Revision 1 Page 11 cf 38 Technical Specification Number Discussion I 3.6 Specifications 3.6.1 through 3.6.8 provide containment integrity requirements to ensure that radiological releases to the environment following an accident inside the Reactor Building are maintained within 10 CFR 100 limits. Since there are no postulated accidents that could occur inside containment in the LTDM whose consequences could come anywhere near the 10 CFR 100 limits, these Specifications are not included in the LTDTS. 3.7 Specifications 3.7.1 through 3.7.4 provide Operability requirements for the normal and emergency power sources. As previously addressed in this submittal (see pages 3 and 4 of this Safety Analysis Report (SAR)), a significant amount of time is available to take corrective action to restore offsite power in the unlikely event of a LOOP. Therefore, an emergency power supply is not required, and these Specifications are not needed to ensure power is provided in a timely manner to support equipment in the LTDM. These Specifications are not included in the LT.0TS. 3.8 Specification 3.8 provides requirements for fuel handling operations in the Spent fuel Building. The Specification 3.8.1 radiation monitor associated with the Spent Fuel Building is included in the LTDTS as Specification D3/4.4. The Applicability of the new Specification has been increased to "whenever irradiated fuel assemblies are in the spent fuel pool" to reflect the importance of monitoring the SFP Building in the LTDM. The Action statement for Specification D3/4.4 is consistent with the requirements of current Specification 3.8.1, except that a reporting requirement has been added for the inoperability of fixed radiation monitoring equipment. The radiation monitors associated with the Reactor Building are not included in the LTDTS since no fuel handling in the reactor building is allowed under the LTDTS. Specifications 3.8.2 through 3.8.11 and 3.8.13 are associated with fuel movement in the Reactor Building, are not applicable in the LTDM, and are not included in the LTDTS. The restriction on movement of loads over irradiated fuel I assemblies in the SFP (Specification 3.8.12) is modified and included in the LTDTS as specification D3/4.3. The load limit chosen (1700 pounds) ensures that no loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool are moved over irradiated fuel assemblies. This is consistent with the assumptions made in the accident analysis for a dropped fuel assembly, and assures in the event such a load was dropped that (1) the activity released would be limited to the activity contained in a single fuel assembly, and (2) any possible distortion of fuel in the SFP will not result in the formation of J a critical array. l

b License Change Safety Analysis Log No. 1091, Revision i Proposed Amendment No. 182, Revision 1 Page 12 cf 38 Technical Specification Number D_iscuts10D 3.9 Specification 3.9.1 is included in the LTDTS as Specification D3/4.2, except, a generic alternate cooling method is referenced instead of specifying one train of the Decay Heat Removal System (DHRS). The District is considering the installation of a new SFC system that would use a single cooling loop and a water to air radiator. Either this new system or one train of the DHRS could provide an alternate means of cooling the SFP. It is unnecessary to restrict the options of the District by designating a specific system as the backup SFC system as is done in Specification 3.9.1. Specifications 3.9.2 and 3.9.3 provide restrictions on the use of DHR as a SFP cooling source when DHR is required for RCS protection. In the LTDM the DHR system is not required for protection of-the reactor. Therefore, these Specifications are not included in the LTDTS. Specification 3.9.4, SFP temperature limit, is modified to remove the reactor operation limitations and include temperature limits with associated Actions to ensure adequate SFP temperature control. A SFP temperature Specification is included in the LTDTS as D3/4.2. Specification 3.9.5, SFP water level, is retained in the LTDTS and modified to include an additional SFP surface radiation surveillance requirement when the SFP water level is below 37 feet. The 2.5 mrem /hr limitation is addressed in USAR Section 9.8. These requirements are included in the LTDTS as Specification 03/4.1. 3.10 Specification 3.10 provides a limit for the amount of Iodine-131 in the secondary side of the steam generators. Since no residual Iodine-131 exists in the secondary plant from previous reactor operation and there is no new production of Iodine-131 in the LTDM, this Specification is not included in the LTDTS. 3.11 Specification 3.11 provides limits on the use of the Reactor Building Polar Crane and Auxiliary Holst. Since there is no fuel in the reactor vessel, the safety consequences of a dropped load in the Reactor Building are negligible. This Specification is not required to help enst.re the protection of the public's health and safety; therefore, this Specification is not included in the LTDTS. /

F License Change Safety Analysis Ltg No. 1091, Revision 1 Proposed Amendment N3. 182, RevisiCn 1 Page.13 cf 38 Technical. Specification Number-Discussion 3.12 Specification 3.12 provides testing requirements for snubbers on safety related systems. Only one snubber remains in the LTDM that is considered safoty related. This snubber is located on the SFC system. This snubber will be maintained under a testing and inspection program consistent with the current Technical Specification required program. The maintenance of this snubber is required to ensure the SFC is capable'of performing its intended function when required by LTDTS Specification D3/4.2, SFP Temperature. With only one snubber involved, there is no justification for a detailed Snubber Technical Specification. The maintenance and testing will be performed to ensure continued acceptable performance of the SFC snubber and will be administratively controlled. by plant-procedures. 3.13 Specifications 3.13.1, 3.13.3 and 3.13.4 provide requirements for the Control Room / Technical Support Center (CR/TSC) and Reactor Building Purge Exhaust filter systems when containment integrity is required and reactor operation is intended. Containment integrity is not required and reactor operation is not allowed in the LTDM. Maintaining CR/TSC Emergency Filtering System Technical Specifications is not required in the LTDM since the source term is_ low enough, due to fission gas decay, that a fuel handling accident will not result in exposure to Control Room personnel anywhere'near allowable habitability levels. The chlorine detector requirements (CR/TSC Emergency Filtering System operability per Specification 3.13, Table 4.1-1, Item 44c, and Specification 4.10.10.2) are required as part of the CR/TSC habitability requirements for plant operations. When the plant is shut down, the CR/TSC Emergency Filtering system is not required to be OPERABLE and the CR/TSC habitability requirements are not applicable. Therefore, in the LTDM the chlorine detector requirements are not applicable and are not included in_ the LTDTS. Specifications 3.13.1 and 3.13.4 address the Operability of the Reactor Building Purge Exhaust Filter System in relation to continued reactor operation and containment integrity. In the LTDM, reactor operation is not allowed and containment integrity l is not required; therefore, these Specification requirements are not included in the LTDTS. The requirements for operating the Reactor Building Purge Exhaust Filter System as a Gaseous Radwaste Treatment system is addressed in Specification 3.18.4. Specification 3.18.4 is being removed from the Technical Specifications per Generic Letter 89-01 and its requirements relocated to the Offsite Dose Calculation Manual (0DCM). l I

24 License Change Safety Analysis < Log N3.1091, Revision 1 Proposed Amendment No.-182, Revisitn 1 Page 14 cf 38 Technical-Specification i Number Discussion 3.13 Specification 3.13.2 provides the requirements for the Operability (Cont.) of the Auxiliary and Spent Fuel Building Filter System when the spent fuel in the SFP has decayed less than 30 days. Since the l spent fuel has decayed for greater than 30 days, this i Specification is not applicable and is not included in the LTDTS. !^ The requirements for operating the Auxiliary and Spent Fuel Building Filter System in the LTDM is associated with the design objectives of 10 CFR 50, Appendix I and is addressed in - [ Specification 3.18.4. The Ventilation Exhaust Treatment System [ portion of Specification 3.18.4 is removed from the Technical Specifications per. Generic Letter 89-01 and relocated to.the ODCM (see the evaluation for Specification 3.18.4 on page 15 of this SAR). 3.14 Specifications 3.14.1 through 3.14.6.2 are Fire Protection l Technical Specifications that were justified for removal pursuant to Generic Letter 88-12 in the District's submittal of Proposed e Amendment No. 180 (PA-180) and Supplement 1 to PA-180. The following list of Specifications have been removed from the Technical Specifications pursuant to the requirements of Generic Letter 89-01 and placed in the ODCH as they currently exist, except as noted in the following evaluations. A revised ODCM that incorporates the existing Radiological Effluent Technical Specifications (RETS) requirements applicable in the LTDM is included as Attachment IV to the PA-182, Revision i submittal package for NRC review. 3.15 Radioactive Liquid Effluent Monitoring Instrumentation 3.16 Radioactive Gaseous Effluent Monitoring Instrumentation 3.17.1 Liquid Effluents Concentration 3.17.2 Liquid Effluents Dose 3.17.4 Liquid Effluent Radwaste Treatment 3.18.2 Gaseous Effluents Dose-Noble Gases 3.18.3 Gaseous Effluents Dose-Iodine-131, Iodine-133, Tritium and Radioactive Materials in Particulate Form 3.18.4 Gaseous Radwaste Treatment (Ventilation Exhaust Treatment Systems only) 3.25 Fuel Cycle Dose Technical Specification Number Discussion 3.17.3 This Specification provides radioactive material content limits for certain outdoor liquid holdup tanks, is retained in the Technical Specifications as required by Generic Letter 89-01, and is included in the LTDTS as Specification D3/4.6. j

1 License Change' Safety Analysis. ' Log N3.1091, Revisitn 1 Proposed Amendment _No. 182. Revisitn 1 Page 15 cf 38 l - Specification Technical Number Discussion 3.18 i The Site Boundary for Gaseous Effluents (LTDTS Figure D5.1-2) will be'used in lieu of the Exclusion Area currently used for a evaluating the dose rate limits-of Specification 3.18.1.. This ?~y change does not affect the current dose calculation methods My specified in the ODCH for gaseous effluents, is consistent with the other gaseous effluent Technical Specifications (3.18.2, 3, and 4), and is consistent with NUREG-0133, Preparation Of Radiological Effluent Technical Specifications for Nuclear Power Plants. Specification 3.18.1 is relocated to the ODCM per Generic Letter 89-01. 3.18.4 The Ventilation Exhaust Treatment Systems (VETS) operating requirements stated in Specification 3.18.4 are relocated to the ODCH in accordance with Generic Letter 89-01.

However, Specification 3.18,4, Gaseous Radwaste Treatment, addresses both the Haste Gas System (HGS) and the VETS.

In the LTDM, the HGS is not required to perform-its intended function or meet the design objectives of 10 CFR 50, Appendix I; therefore, operating requirements for the HGS are not included in the LTDTS. The HGS is designed to collect and hold for eventual release fission product gases generated in the Reactor Coolant System (RCS).during reactor operation. This design meets the design objectives of 10 CFR 50, Appendix I for an operating plant. Now ( that the reactor has been defueled and the RCS depressurized and vented' to atmosphere for an extended period of time, essentially all tne fission product gases contained in the RCS and collected -in-the HGS have'been released through the HGS. Hith negligible amounts of radioactive gases available for discharge through the HGS in the LTDM, the HGS is not required to be used as a Gaseous Radwaste Treatment _(GRT) system to help ensure to 10 CFR 50, [ Appendix I guidelines are met in the LTDM. This Appendix I design objective is met through the use of the VETS as GRT systems, because the composition of gaseous effluent in the LTDM will be almost exclusively particulate matter with no Iodine. Therefore, in the LTDM the HGS is not required to collect and hold gases for Appendix I considerations, and the GRT requirements for the HGS are not included in the LTDTS or the ODCM. For further discussion on the WGS, see the evaluation for Specifications 3.18.5 and 3.24. t j

a,,i' l-i.e.......-.-....... License Change Safety' Analysis-. Log No.1091, Revision 1 Pr:pos d Amendment ND.182, Revision 1 Page 16 of-38 Technical Specification-Number Discussion 3.18.5 Gas Storage Tanks 3.24 Explosive Gas Mixture Specifications 3.18.5 and 3.24 provide radioactivity content and explosive gas mixture limits for the Haste Gas Decay Tanks (HGDTs), which are components of the HGS. The HGDTs are used to store fission product gases and other off-gases from the Reactor Coolant System (RCS) produced during reactor operation and removed from the RCS by the Makeup and Purification System. During normal . plant operation, hydrogen is used as the cover gas in the Makeup Tank to provide a means of oxygen scavenging in the RCS. Now that the reactor has been defueled and the RCS depressurized and vented to atmosphere for-an extended period of time, the fission product gases and hydrogen normally entrained in the RCS have been released through the HGDTs. Any remaining radioactive or other gases normally collected in the HGDTs will be discharged through the Auxiliary Building Stack Ventilation System without being held up in the WGDTs. Thus, no potential exists for accumulating an explosive gas mixture in the WGDTs. Also, the quantity of radioactive gases available to the HGS in the LTDM is minimal. The design basis for the activity content limit on the HGDTs assumes there is fuel in the reactor and an excessively high RCS specific activity. This design basis condition is not applicable in the LTDH. Therefore, it is not necessary to hold up radionuclides in the HGDTs or monitor the concentration of oxygen in the HGDTs during the LTDM, and the requirements of these Specifications are not applicable and are not included in the LTDTS. 3.21 This Specification provides requirements for Solid Radioactive Hastes and has been removed from the Technical Specifications pursuant to the requirements of Generic Letter 89-01 and placed in the Process Control Program (PCP). The PCP was included as part of the original PA-182 submittal package for NRC review. 3.22 Radiological Environmental Monitoring 3.23 Land Use Census Specifications 3.22 and 3.23 have been removed from the Technical Specifications pursuant to the requirements of Generic Letter 89-01 and placed in the Radiological Environmental Monitoring Program (REMP) Manual. The REMP Manual was included as part of the original PA-182 submittal package for NRC review. 3.26 This Specification provides Interlaboratory Comparison Program requirements that are applicable to the REMP. This requirement has been removed from the Technical Specifications pursuant to Generic Letter 89-01 and placed in the REMP Manual. l

License Change Safety Analysis Log No.1091, Revision 1 - Proposed Amendment N).-182, Revisitn 1 Page 17 cf 38 Technical Specification Number Discussion 3.27 Specification 3.27 provides requirements for the Nuclear Service Electrical Building (NSEB) Emergency Heating Ventilation and Air Conditioning (HVAC) System. This system provides a backup emergency HVAC system for the NSEB and the electrical equipment contained within the building. The requirements of this Specification are applicable in the Heatup through Power Operations modes, and are not applicable in the LTDM. With the plant shutdown and the reactor defueled, the heat load in the NSEB is significantly reduced since most of the equipment supplied from buses in this building is deenergized. A significant period of time is available to allow restoration of a HVAC system should the operating system become unavailable. A Specification for these systems is not included in the LTDTS. 3.28 This Specification provides requirements for the TDI Diesel Generator Control Room Essential Ventilation System and is applicable in the Heatup through Power Operations modes. The emergency power sources are not required to be maintained OPERABLE in the current plant condition. Based on the LOOP analysis and the requirements for the current plant condition, the requirements of Specification 3.28 are not applicable in the LTDM and are not included in the LTDTS. 3.29 This Specification provides requirements for the Meteorological Monitoring Instrumentation. The data collected-from this instrumentation is currently used in dose calculations. The extended decay time of the fuel (327 days as of April 30, 1990) has eliminated the possibility of an offsite dose to a member of the public that would require protective action. -Also, the activity level of-normal gaseous effluent releases from the site 1 has decreased to the point where multi-year average meteorological values for offsite dose calculations can be used without affecting Rancho Seco's ability to maintain within the 10 CFR 50, Appendix I dose guidelines. An eleven year operating baseline of meteorological data is available for determining multi-year averages. In the LTDM, the Meteorological Monitoring Instrumentation required by the Emergency Plan will be administrative 1y controlled and maintained through plant procedures to ensure this instrumentation will function as required. Therefore, this Specification is not included in the LTDTS. 3.30 .This Specification provides requirements for the Hydrogen Recombiners which control hydrogen gas inventory in the H containment following a LOCA. The Hydrogen Recombiners are l required to be Operable when the reactor is subcritical by less than 1 percent ak/k. Since the reactor is defueled the Hydrogen j l Recombiners are not required in the LTDM and are not included in l the LTDTS. l [ l l

License Change Safety Analysis Log ND. 1091, Revisitn 1 Proposed Amendment ND, 182, Revision 1 Page 18 of 38 L Technical Specification Number Discussion 4.0.1 This Specification is included in the LTDTS as D4.0.1.. Editorial changes are made that omit reference to operational modes. 4.0.2 This Specification has been modified for applicability in the LTDM and is included in the LTDTS as Specification D4.0.2. I 4.0.3 This Specification provides restrictions on entry into an operational mode. Since the LTDTS apply only in the LTDM, this Specification does not apply and is not included in the LTDTS. q 4.0.4 This Specification provides an extension for certain surveillances l to allow their performance during the Cycle 8 refueling outage. This Specification is not applicable in the LTDM and is not included in the LTDTS. ,j 1 4.1 This Specification provides surveillance and testing requirements for instrumentation associated with RPS, SFAS, Process In:;trumentation, and Emergency Shutdown Instrumentation. Most of this instrumentation is required for reactor operation and transient monitoring, and accident identification and mitigation for an operating nuclear power plant, which is not applicable in the LTDM.- But, several items from Table 4.1-1 are applicable in the LTDM. Table 4.1-1. Items 44a, 44b, 45, and 63 are dispositioned as l follows: a) Surveillance and testing requirements for those radiation monitors required by the Standard RETS (NUREG-0472) and associated with radioactive effluent control and monitoring have been relocated to the ODCM in accordance with the requirements of Generic Letter 89-01. (See the evaluation for Specifications 3.15, 3.16, 3.17 and 3.18 on pages 14 and 15 of this SAR.) b) The SFP area radiation monitor is included in the LTDTS as Specification D3/4.4. i l c) The remaining radiation monitors are used for general area and L process monitoring, are not included in the Standard Technical Specifications, are controlled by administrative and technical procedures, and are therefore not included in the LTDTS. l l l l h

g License Change Safety' Analysis Log N3, 1091, Revisicn 1 Prtposed Amendment N3. 182, Revision 1 Page 19 of 38 Technical Specification Number Discussion d) The surveillance requirements'for Item 45 of Table 4.1-1, Emergency Plant Radiation Instruments, are not included in the LTDTS because-these requirements are included in an - implementing procedure for the approved Emergency Plan. Changes to this procedure would-require both a 10 CFR 50.59 and 10 CFR 50.54(q) review to determine if an Unreviewed Safety Question or a reduction in the effectiveness of the Emergency Plan would result. Also, the Nuclear Safety Review and Audit Committee and the AGM, Nuclear is required by the LTDTS (Specifications 06.5.1.7d and D6.5.3a, respectively) to review all Emergency Plan implementing procedure changes. Therefore, the surveillance requirements specified in Table 4.1-1, Item 45 are not included in the LTDTS. Table 4.1-1, Item 44c, Chlorine Detector, is not applicable in the LTDM and is not included in the LTDTS. See the evaluation for Specification 3.13 (page 13 of this SAR). Table 4.1-1 Item 46 Environmental Air Monitors, has been relocated to the REMP in accordance with Generic Letter 89-01. Table 4.1-1, Item 47, Strong Motion Accelerometer, is used to provide information for plant shutdown and cooldown following a Safe Shutdown Earthquake or Design Basis Earthquake. Since the plant is already shut down, the accelerometer would only provide information that is also available from State or Federal Agencies. Therefore, this instrument is not included in the LTDTS

since this type of information is available from other sources.

Table 4.1-1, Item 51, SFP Level, is included in LTDTS Specification D3/4.1. The RPS and SFAS related surveillance requirements listed in Table 4.1-1 (Items 1 through 27) apply to an operating nuclear power plant, are not applicable in the LTDM, and are not included in the LTDTS. The reactor operation, transient, and accident monitoring instrumentation and Emergency Core Cooling System process instrumentation surveillance requirements addressed in Table 4.1-1 (Items 28 through 43, 49, 50, 52 through 62, 64 through 56, 58 through 76, and 84 through 90) are not applicable in the LTDM and are not included in the LTDTS (No surveillances are associated with Items 48 and 77 through 81), 4 ,] ~h

r i License Change Safety Analysis ~ Log No. 1091, Revision-l' - Pr@ posed Amendment No. 182, Revision 1 Page 20 of 38 Technical-Specification Number -Djscussion

4. l' Table 4.1-1, Item 57, Voltage Protection, is not required because (Cont.)

if degraded voltage or a LOOP is experienced, several hours are available to regain-electric' power and re-establish SFC (see the evaluation on page 3 of this SAR). The emergency power sources are not. required in the LTDM. Therefore, Item 57 is not included in the LTDTS. Table 4.1-1, Items 82 and 83, Spray Pond Hater Temperature and Level, is not included in the LTDTS based on the LOOP analysis and the amount of time available to restore SFC. Table 4.1-2, Items 1 through 4, 6, 7, 10, and 12 through 19 are surveillances that are required for a plant with fuel in the reactor or a plant that intends to operate. In the LTDM no fuel is-allowed in the reactor building and no intent to operate the plant exists. Therefore, these surveillances are not included in the LTDTS. Table 4.1-2 Item 5, Refueling System Interlocks, is included in LTDTS Specification 03/4.3 to require testing of the Spent Fuel Building fuel handling system interlocks only. The Applicability Statement for this surveillance ensures completion of Functional Testing within 7 days prior to fuel handling. Table 4.1-2, Item 8, Charcoal and High Efficiency Filters, is not included in the LTDTS. Because the plant has been shut down since June 1989, iodine levels in the plant have decayed away to essentially zero. Therefore, charcoal filtration units are not required in the LTDM. Also, the radioactive gaseous effluent requirements which address the operation of the HEPA Filters (Spec'Ifications 3.18.4 and 4.22.4) have been relocated to the ODCM in accordance with Generic Letter 89-01. For further discussions on the requirements of the charcoal and HEPA Filters in the LTDM, see the evaluations for Specifications 3.13, 3.18.4, and 3.18.5 (pages-13 through 16 of this SAR). The Table 4.1-2, Item 9, Fire Pumps and Power Supplies, is relocated to the Fire Protection Plan pursuant to the requirements of Generic Letter 88-12. Relocation of the fire protection requirements is addressed in Proposed Amendment No.180.

1 o License ~ Change' Safety Analysis Log No. 1091, Revisicn 1 Proposed Amendment'NO 182, Revisien 1 Page 21 cf 38 . Technical Specification Number Discussion 4.1 Table 4.1-2,.~ Item 11, SFC System, requires functional testing of (Cont.). the SFC system each refueling interval prior to fuel handling.- This requirement.is met by continuously maintaining the SFP temperature below 140 degrees Fahrenheit. cThe intent of the Specification is to. test the SFC system prior to the addition of more spent fuel to the SFP. Because no more Spent Fuel will'be s 'added to the SFP in the LTDM, the heat load in the SFP will. continue to decrease from a calculated value of 3.6 million BTU / hour as of January 3,1990 to 2.8.million BTU / hour on June 7, 1990. This compares to a design capacity of the SFC of 8.76 million BTU / hour. Extraordinary system degradation would have to occur for the system to not meet its intended function. Therefore, the Functional Test will consist of verifying normal operation of the SFC System'and maintaining the SFP temperature .below 140*F. Implementation of corrective actions if significant SFC performance degradation is detected will be required and addressed in plant administrative procedures. Table 4.1-3, Items 2,4,-and 6, Maintenance of a Minimum Boron Concentration in the Borated Hater Storage Tank, SFP, and Concentrated Boric Acid Tank, are not required in the LTDM since borated water is not needed to maintain adequate shutdown margin in the SFP (see evaluation on page 3 and 4 of this SAR). Therefore, these requirements are not included in the LTDTS. Table 4.1-3, Items 1, 3, 5, 7 and 8 are applicable only for an operating plant with fuel in the reactor. In the LTDM these surveillance requirements are not applicable and are not included in the LTDTS. 4.2 Specifications 4.2.2.1 through 4.2.2.5 provide inspection, examination, and testing requirements for systems as called out in ASME Section XI for Code Class 1, 2, and 3 components. The inspections identified verify the integrity over time of high 1 temperature, high pressure systems. The systems remaining at Rancho Seco are-at relatively low pressure and temperature when compared to their' design limits. Therefore, erosion and corrosion expected at high temperature and pressure would not be expected. ASME Section XI, section IHA-2400(c)-provides for the deferral of inspections due to extended plant outages. Being in the LTDM constitutes an extended plant outage. While the plant is in the LTDM, these inspection and examination requirements will not be performed and thus are not included in the LTDTS.

LicenseLChange Safety Analysis-Log No.1091', Revisitn 1 Proposed Amendment N). 182. Revision 1 Page 22 cf 38 -Technical -Speciftcation Number Discussion 4.2-The testing requirements for pumps and valves are established to (Cont.') verify equipment is capable of meeting their design requirements. The systems remaining in service at Rancho Seco are experiencing a continual decrease in required capability as a function of time due.to the decay of the spent fuel. The pumps and valves required .to' remain in service will be maintained under the District's preventive maintenance program which includes vibration, pressure drop, and flow monitoring for rotating equipment and systenis in use in the LTDM to identify problems prior to failures. This program is similar to-that which would be implemented under Section XI. 'These requirements will be administratively controlled and are not included in the LTDTS. 4.2.3 This requirement is the surveillance for Specification 3.1.6, which is only applicable above Cold Shutdown. Monitoring the RCS for leakage is unnecessary in the LTDM because the reactor is defueled. This surveillance is not included in the LTDTS. See the evaluation for Specification 3.1 on page 8 of this SAR. 4.3 The verification of RCS integrity prior to returning to criticality is not applicable in the LTDM and is not included in the LTDTS. 4.4 Specifications 4.4.1.1 through 4.4.2.6 provide testing, inspection, and-inservice surveillance requirements which monitor p the structural and leakage integrity of the Reactor Building. In the LTDM, there are no credible accidents that require Reactor Building integrity. Maintaining Reactor Building integrity as specified in these surveillance requirements is no longer required -to provide reasonable assurance that activities conducted at the facility will not result in undue risk to the health and safety of the public. Therefore, these surveillance requirements are not required in the LTDM and are not included in the LTDTS. L 4.5 Specifications 4.5.1 through 4.5.3 provide testing requirements to verify proper operation of the Emergency Core Cooling and Reactor Building Spray systems. With the reactor defueled, the need for these systems no longer exists. Therefore, these surveillance requirements are not included in the LTDTS. The DHR system leakage testing requirement in Specification 4.5.3 is predicated on assuring that the system will operate to cool the reactor for approximately 200 days and will minimize the dose due to leakage from this system following a loss of coolant accident (LOCA). In the LTDM, a LOCA as defined in the USAR is not credible; therefore, this requirement is not included in the LTDTS. l u e

a License Change ' Safety Analysis-Log ND. 1091, Revisitn 1 Proposed Amendment N3, 182,.Revisicn 1 Page 23 cf 38 Technical-Specification 4 Number-Discussion 4.6 The emergency power systems are not required when the plant is' . shut down,~thus the surveillances for.these systems are not required in the.LTDM and are not included -in the LTDTS. See the evaluation for the LOOP event and Specification 3.7 on pages 3 and 11 of this_SAR, respectively. 4.7 Specifications 4.7.1.1 through 4.7.2.3 provide testing -requirements for the control rods. Since the_ reactor is defueled, these Specifications are not applicable in the LTDM and are not included in the LTDTS. 4.8 Specifications 4.8.1 through 4.8.5 provide testing requirements for the Auxiliary Feedwater System. Since the reactor is defueled, there is no need to provide a feedwater supply for the steam generators. Therefore, these Specifications are not included in the LTDTS. 4.9 Specification 4.9 requires evaluation of reactivity anomalies that occur during plant operation. Since the reactor is defueled this -Specification is not applicable in the LTDM and is not included in the LTDTS. 4.10 This surveillance requirement is not applicable in the LTDM and is not included in the LTDTS. See the evaluation for Specification 3.13 on pages 13 and 14 of this SAR. 4.11 See the evaluation for Specifications 3.13.1, 3.13.4, 3.18.4, and Table 4.1-2, Item 8 on pages 13, 14, 15, and 20 of this SAR. 4.12 See the evaluation for Specifications 3.13.2, 3.18.4, and Table 4.1-2, Item 8 on pages 14,15, and 20 of this SAR. 4.13 This Specification provides inspection requirements for specific-high energy _ lines outside the Reactor Building. Since the high energy systems (Main Feedwater and Main Steam) associated with this inspection requirement are not in service in the LTDM, this Specification is not applicable and is not be included in the LTDTS. 4.14 See the evaluation for Specification 3.12 on page 13 of this SAR. 4.15 The requirements of this Specification are rewritten to more precisely reflect the requirements of 10 CFR 39.35, Leak Testing of Sealed Sources. Significant editorial changes are made to clarify the 10 CFR 39.35 requirements. This Specification is incluaed in the LTDTS as D3/4.7. 4.16 This Specification was previously deleted. )

' License Change Safety Analysis Log No. -1091, Revision 1 Proposed Amendment No. 182, Revision 1 Page 24 of 38-Technical Specification Number Discussion 4.17 This Specification provides testing requirements for the Steam Generators. Since the Steam Generators are not in service in the LTDM, these requirements are not included in the LTDTS. 4.18 See the evaluation-for Specification 3.14 on page 14 of this SAR. 4.19 See the evaluation for Specification 3.15 on page 14 of this SAR. 4.20 See the evaluation for Specification 3.16 on page 14 of this SAR. 4.21.1 See the evaluation for Specification 3.17.1 on page 14 of this SAR. 4.21.2 See the evaluation for Specification 3.17.2 on page 14 of this SAR. ,{ 4.21.3 See the evaluation for Specification 3.17.3 on page 14 of this SAR. l 4.21.4 See the evaluation for Specification 3.17.4 on page 14 of this SAR. 4.22.1 See the evaluation for Specification 3.18.1 on page 15 of this SAR. l 4.22.2 See the. evaluation for Specification 3.18.2 on page 14 of this SAR. j 4.22.3 See the evaluation for Specification 3.18.3 on page 14 of this SAR.- 4.22.4 See the evaluation for Specification 3.18.4 on page 15 of this SAR. 4.22.5 See the evaluation for Specification 3.18.5 on page 16 of this SAR. 4.23 This Specification was previously deleted, i 4.24 This Specification was previously deleted. 4.25 See the evaluation for Specification 3.21 on page 16 of this SAR. 4.26 See the evaluation for Specification 3.22 on page 16 of this SAR. 4.27 See the evaluation for Specification 3.23 on page 16 of this SAR. 4.28 See the evaluation for Specification 3.24 on page 16 of this SAR. 4.29 See the evaluation for Specification 3.25 on page 14 of this SAR. 4.30 See the evaluation for Specification 3.26 on page 16 of this SAR. I 4.31 See the evaluation for Specification 3.27 on page 17 of this SAR. L

License Change Safety Analysis Log N3. 1091, Revisi;n 1-Proposed Amendment No. 182, Revision 1 Page 25 of 38 Technical Specification . Number Discussion 4.32 See the evaluation for Specification 3.28 on page 17 of this SAR. 4.33 This Specification was previously deleted. 4.34 See the evaluation for Specification 3.29 on page 17 of this SAR. 4.35 See the evaluation for Specification 3.30 on page 17 of this SAR. J 5.1 Figures 5.1-1 and 5.1-2, Exclusion Area (EA) and Low Population Zone (LPZ), respectively, are combined into a single Figure (LTDTS Figure D5.1-1) titled Emergency Planning Zone (EPZ). This new 1 boundary is established in accordance with the defueled condition Emergency Plan. The EPZ is reduced to the industrial area boundary outlined in LTDTS Figure 05.1-1. The size of the EPZ boundary.is based on the minimal consequences that result from the few accidents considered credible in the LTDH. At this boundary, calculated radioactive exposures do not exceed the Protective Action Guidelines of EPA 520/1-75-001-A, January 1990, Manual of Protective Actions for Nuclear Incidents, for the worst caso defueled condition design basis accident (the dropped fuel assembly. accident). The EA and LPZ are acceptability criteria used by the NRC to determine suitability of proposed reactor sites in relation to 10 CFR 100 release limits, are not applicable in the LTDM, because the dose consequence of the dropped fuel assembly accident-in-the LTDM is an extremely small fraction of the 10 CFR 100 limits (see the-evaluation on page 2 of this SAR) and are replaced by the EPZ. The titles of Figures 5.1-3 and 5.1-4 are modified in accordance with the Standard Radiological Effluent Technical Specifications (NUREG-0472) to clarify the application and significance of the two effluent boundaries. Also, the two figures are renumbered DS.1-2 and D5.1-3 and included in the LTDTS. The gaseous effluent boundary (LTDTS Figure D5.1-2) applies to gaseous effluents released from the site under normal conditions and is used to evaluate both 10 CFR'20 and 10 CFR 50, Appendix I compliance. The liquid effluent boundary (LTDTS Figure D5.1-3) identifies the boundary at which radiological liquid effluent actually leaves the industrial area. The dose accountability points for meeting the 10 CFR 50, Appendix I guidelines remains at the A and B Regenerant Hold-Up Tanks, and the concentration accountability points for determining 10 CFR 20, Appendix B compliance remains at the Retention Basins. Under accident conditions the EPZ boundary is used to evaluate the potential consequences of an accident and the corresponding response and notification actions required in accordance with the defueled condition Emergency Plan. The EPZ boundary is the same as the liquid effluent boundary, and more conservative than the gaseous effluent boundary.

License Change Safety Analysis. Log No. 1091, Revision 1 Proposed Amendment Na 182, Revisitn 1 Page 26 cf 38 i Technical Specification Number Discussion 5.2 The two systems addressed in this Specification, Reactor Building and Reactor Building Isolation, are not required to be OPERABLE in the LTDM, thus this Specification is not included in the LTDTS. 5.3 This section describes the Reactor. Since the Reactor is defueled, this Specification is no longer applicable and is not included in the LTDTS. 3 5.4 The descriptions of spent fuel storage in the Spent Fuel Storage Building have been included in the LTDTS as Specification D5.2. The remainder of this Specification addresses new fuel storage, is not applicable in the-LTDM, and is not included in the LTDTS. 6.1 The responsibilities specified for facility management are redefined for the LTDM and included as Section D6.1 of the LTDTS. The titles have been~ changed to reflect the present organization. Directives are in place to delineate responsibility for the current Technical Specification duties. 6.2.1 This Specification has been changed to reflect new position titles, the LTDM, and is included as LTDTS D6.2.1. 6.2.2 Specifications 6.2.2a and 6.2.2h are combined and included in the LTDTS as Specification D6.2.2a. Specification D6.2.2a references the minimum operator shift crew requirements of Table D6.2-l'and modifies the qualification requirement of the individual who.

supervises the Shift Supervisors. The title Shift Operations

' Superintendent is removed and replaced with a general description of the position. The modified qualification requirement for the individual who directly supervises the Shift Supervisors is consistent with the attached training programs and the LTDM. Table 6.2-1 Shift Crew Personnel and License Requirements, is modified to reflect the LTDM and is included in the LTDTS as Table D6.2-1, Minimum Shift Crew Requirements. As indicated in Table e 06.2-1, the qualification requirements for a Shift Supervisor is no longer a NRC licensed senior operator, but rather a Certified Fuel Handler (CFH) pursuant to the requirements contained in Attachment III, Certified-Fuel Handler Training Programs. In the LTDM, the requirement for NRC licensed operators is removed. With the reactor defueled to the spent fuel pool, plant closure activities in progress, and no intention by the District to operate the plant, the District has determined that the requirements of 10 CFR 55 for licensed operators and senior licensed operators are not applicable. Training programs are provided for NRC review and approval (Attachment III), pursuant to l the guidance of 10 CFR 72.192 and the NRC's Nuclear Materials Safety and Safeguards General Plan for Operator Certification under 10 CFR 72.

License Change Safety Analysis Log N3. 1091, Revision'1 Proposed Amendment N3.182, Revisicn 1 Page 27 cf 38 -Technical Specification i Number Discussion 6.2.2 The minimum shift crew requirements for the LTDM meet the number (Cont.) of personnel required by 10 CFR 50.54(m) for a facility with a single, non-operating unit, but exclude the NRC licensed operator requirements. A current plant operator can qualify as a Shift Supervisor by becoming a CFH in accordance with the-training programs provided for NRC approval. The Shift Supervisor continues to be charged with the control room command function. The training and qualification requirements contained in the new training programs for the. Shift Supervisor.are commensurate with the Shift Supervisor's fuel handling, control room command function, and supervisory responsibilities in the LTDM. The change from NRC licensed operators to District certified fuel handlers is commensurate with the change of Rancho Seco from an operating reactor facility to a permanently and completely defueled facility with all fuel stored in the spent fuel pool. Pursuant to 10 CFR 55.2(a) individuals who m nipulate the " controls" of a utilization facility, licensed pursuant to 10 CFR 50, must be NRC licensed operators. " Controls" are defined in 10 CFR 55.4 as " apparatus and mechanisms the manipulation of which directly affects the reactivity or power level of the reactor". With all fuel stored in the spent fuel pool,-there are no controls that can be manipulated which directly or indirectly affect the-reactivity or-power level of the reactor. Therefore, the District has concluded that the requirements of 10 CFR 55 do not apply to Rancho Seco in the long term defueled condition. A formal review of the change in the training program was performed pursuant to the requirements of 10 CFR 50.59. This review determined that the change does not involve an unreviewed safety question. Currently, approximately 25 members of the plant staff hold a valid NRC senior operator license. By virtue of the training and qualification requirements met to obtain this license, these individuals meet the requirements for a CFH. The appropriate number of additional personnel will be trained and qualified as a CFH in accordance with the approved training programs.to ensure the minimum shift crew requirements of Table 06.2-1 for the LTDM are met. Specifications 6.2.2b through 6.2.2d are applicable only when fuel is in the reactor; therefore, these requirements are not applicable in the LTDM and are not included in the LTDTS. Specification 6.2.2', applicable during core alterations, is replaced by LTD'S Specification D6.2.2b which applies to spent fuel handling in the LTDM.

a 1 License' Change Safety _ Analysis Page 28 cf 38 j Proposed Amendment N3.182 Revision 1 Log NL 1091, Revision 1 1 ' Technical Specification __ Number Discussion 6.2.2 Specification 6.2.2.f defines the Fire Brigade requirements and is (Cont.) removed from the Technical Specifications per Generic Letter 88-12 and placed in the Fire Protection Plan as specified--in Proposed-Amendment No. 180 (PA-180). Specification 6.2.2.g requires the existence of administrative procedures that limit overtime for certain personnel who perform safety-related functions. This restriction is tied to plant operation since the postulated accidents posing significant risk to plant personnel and the health and safety of the public. occur during plant operation and not in the LTDM. Therefore, Specificaticn 6.2.2.g is not included in the LTDTS. 6.3.1 This Specification is included in the LTDTS as Specification D6.3, except the requirement for a Shift Technical Advisor (STA) has been deleted since an STA is not required in Cold Shutdown,-and thus, not required in the LTDM. 6.4.1 This Specification has been modified to reference the CFH training programs (Attachment III) and is included in the LTDTS as Specification D6.4. Besides the retraining and replacement training of Certified Fuel Handlers, which is specifically addressed in LTDTS Section D6.4, the remainder of the plant staff will be trained in accordance with plant administrative procedures to maintain a level of staff competency commensurate with a plant in the LTDM. Plant staff qualification requirements will continue to meet the requirements of ANSI N18.1-1971. Because there is no regulatory requirement to have a NRC approved training program for non-licensed staff at an operating nuclear power plant, no requirement for a NRC approved non-licensed staff training program is included for the LTDM in i the LTDTS. 6.4.2 This Specification for Fire Prigade training was removed per Generic Letter 88-12 and is included in the Fire Protection Plan-as specified in PA-180. 6.5 Specifications 6.5.1 and 6.5.2, the Plant Review Committee (PRC) and Management Safety Review Committee (MSRC) requirements, are combined and rewritten into a single Specification, LTDTS D6.5.1, (the Nuclear Safety Review and Audit Committee (NSRAC) requirements), to address the LTDM and the reduced plant activities, review, audit, and oversight requirements associated with a plant that is defueled. Also, the LTDTS Specification for the new NSRAC committee is written to reflect the reduced potential consequences from the limited credible accidents that could result from facility activities conducted in the LTDM.

r;n i License Change Safety Analysis Log No.-1091. Revisicn 1 - Proposed Amendment N3.182, Revisitn 1 Page 29 cf-38 Technical Specification Number Discussion 6.5 Personnel who make up'the NSRAC are required to meet the current l L-(Cont.) . Technical Specification qualification requirements specified for. l the PRC or MSRC. These qualification requirements are-L (1) personnel filling positions in the Nuclear Organization who meet or exceed ANSI N18.1-1971, Sectiol 4.2 or 4.4, or (2) other personnel who meet or exceed ANSI /ANS 3.1-1981, Section 4.7.2. This qualification requirement for NSPAC members is included in the LTDTS as Specification D6.5.1.2. The function, composition, meeting frequency, quorum requirements, review responsibilities, authority, and records requirements for. the NSRAC are a combination of existing PRC and MSRC requirements rewritten such that the NSRAC requirements are commensurate with a I plant in the LTDM. The NSRAC requirements are included in the LTDTS subparts that comprise Specification 06.5.1. LTDTS Specification 06.5.2 is not used. l Specification 6.5.3 is editorially modified to clarify the l applicability of the review requirements for specific procedures, l plans, manuals, and programs and to reflect the current organization. These requirements-are included in LTDTS Specification 06.5.3. The audit requirements of Specification 6.5.4 are modified to i reflect the non-operating status and defueled condition of the plant and to address the activities conducted in the LTDM, and are included in the LTDTS as Specification D6.5.4 with the following changes in audit' frequencies: l^

1. The frequency of audit requirement 6.5.4c is changed to once L

per year from once per six months. This reduction in frequency is consistent with the reduction in plant activity associated-with a plant in the LTDM.

2. The Facility Security Plan audit frequency is changed to yearly from two years'to reflect the audit performed in accordance with 10 CFR 73.55(g)(4).

3. The audit frequency of the REMP and implementing procedures is changed to 2 years from 12 months to be consistent with the standard RETS and the audit frequency for the ODCM and PCP.

Also, as required by Generic Letter 89-01 and the standard RETS, the current annual audit of REMP results is retained. Audit requirements 6.5.41 and 6.5.4j address Fire Protection, are editorially combined into a single Specification, and are included. o L in the LTDTS as Specification D6.5.41. 1 l-l 1

c License Change Safety Analysis ' Log ND. 1091, Revision.1 Proposed' A:nendment N3.182, Revision 1 Page 30 of 38 q Technical Specification Number Di cussion 6.6 This Specification, which addresses Licensee Event Reports, is j combined with Specification 6.9.4, Licensee Event Report, and l included in the LTDTS as Specification D6.9.4; therefore, LTDTS -Section D6.6 is not used. 6.7 This Specification describes the reporting actions required if a Safety Limit, as defined in Technical Specification Section 2 is-violated. Since the Section 2 Safety Limits are not applicable in -the LTDM (see the evaluation for Section 2 on page 8 of this SAR), this Specification is not included in the LTDTS. LTDTS Section D6.7 is not used. 6.8 This Specification has been modified to reflect those procedures, plans, programs, and manuals required for a plant in the LTDM and is included in the LTDTS as Specification 06.8. For the LTDM, the Certified Fuel Handler-training programs and the Quality Assurance program arc added to the list of Procedures, Plans, Manuals, and Programs included in LTDTS Specification D6.8.1. In addition, requirements for the Radioactive Effluent Controls Program and the-i REMP are included in the LTDTS in accordance with Generic Letter 89-01 as Specifications D6.8.3a and D6.8.3b, respectively. As ~previously justified in Proposed Technical Specification Amendment No. 155 and approved-by the NRC in Amendment No. 98, the NRC granted the District a deviation from the Standard RETS requirement (NUREG-0472) relating to the dose projection. limit for operation of the liquid effluent radwaste treatment system. The NRC approved a 31-day dose projection limit of 81/3 percent of the annual 10 CFR 50, Appendix I dose guidelines, on liquid effluent radwaste treatment system operation, instead of the Standard RETS 31-day dose project limit of 2 percent of the annual guidelir,es. l This relief is necessary because Rancho Seco is a dry site, thus making compliance with 10 CFR 50, Appendix I more difficult. Therefore, the 31-day dose projection limit which currently exists in the Technical Specifications is retained for the LTDM and is included in the LTDTS as Specification D6.8.3a.7. - Also, due to the absence of Iodine-131 and Iodine-133 in the plant and the absence of any productior, trochanism for these isotopes in y the LTDM, requirements addressing'these isotopes are excluded from I LTDTS Specification 06.8.3a.10. L 6.9.1 Specifications 6.9.1.1 through 6.9.1.4 provide requirements for submitting various startup related reports. These reporting I requirements are not applicable in the LTDM and are not included in the LTDTS. 6.9.2.1.1 This Specification, Annual Occupational Radiation Exposure Report, is included in the LTDTS as Specification 06.9.1.1. L

License' Change Safety Analysis Log N3. 1091, Revision 1-Proposed' Amendment N3.182, Revisitn 1 Page 31 of 38 Technical Specification Number Discussion 6.9.2.1.2 This Specification Annual Exposure Report, is _ included in the LTDTS as Specification D6.9.1.2. 6.9.2.2 This Specification, Annual Radiological Environmental Operating Report, was modified to meet the requirements of Generic Letter 89 and is included in the'LTDTS as Specification D6.9.1.3. 6.9.2.3 This Specification, Semiannual Radioactive Effluent Release Report, was modified to meet the requirements of Generic Letter 89-01 and is included in the LTDTS as Specification D6.9.2. 6.9.3 This Specification provides the requirements for the contents of the Monthly Report. This report addresses primarily' plant operating statistics and experiences. For the LTDM, the reporting period is changed to annually, which is consistent with the requirements of 10 CFR 50.59(b)(2), and the required content of the report is modified to reflect the significant reduction in plant activities anticipated for a defueled reactor. The Annual Report is included in the LTDTS as Specification D6.9.3. 6.9.4 This Specification provides requirements on reportability pursuant to 10 CFR 50.73. The detailed requirements have been removed and replaced with a reference to meet !0 CFR 50.73 to avoid a conflict with potential future changes to the Code.. Specification 6.6 is included in 6,9.4 to consolidate' tha LER Specifications. See the evaluation for Specification 6.6 on page 30 of this SAR. LER requirements are included in the LTDTS as Specification D6.9.4. 6.9.5 The Special Reports' require d by this Specification are not included in the LTDTS for the following reasons: 6.9.5.A was a one time only report required for 1977. 6.9.5.B,C,D,F, and P are associated with an operating,,iant only. 6.9.5.E is relocated to the Fire Protection Plan in PA-180 pursuant to Generic Letter 88-12. 6.9.5.G,H I,J,K,M,N, and 0 have been removed and relocated per ' Generic Letter 89-01. 6.9.5.L was previously deleted. Therefore, LTDTS Section 06.9.5 is not used. 6.9.6 This Specification, Environmental Reports, is included in the LTDTS as Specification D6.9.6. 1

-License Change Safety Analysis-Log No. -1091, Revision 1 -9 Proposed Amendment No. 182, Revision 1 Page 32 of 38 Technical. Specification-Number Discussion 6.10 The requirements for records retention are included.in the LTDTS as Specification D6.10. The existing record retention requirements are retained in the LTDTS. At the time the facility operating license is modified,- the record retention' requirements-applicable to an operating plant will be modified. Some record . retention requirements are added to address a plant in the LTDM. Additions made to this Specification are described below. D6.10.1j-is added to ensure records and lo'gs of facility activities in the LTDM are retained for at. lease five years. A provision is added to Specification 6.10.21_ Environmental-Qualification Records, to ensure records of (.hanges made to the environmental qualification of equipment while the plant is in the LTDM are retained for the duration of the facility license. This requirement is included in the LTDTS as Specification D6.10.21. A requirement for the retention of records performed for reviews of changes made to the ODCH, REMP HANUAL, and the PCP is added to the LTDTS in accordance with Generic Letter 89-01 as Specification D6.10.20. A requirement for the retention of records related to plant closure activities is included in the LTDTS as Specification D6.10.2p. Specification D6.10.2q is added to ensure records of NSRAC meetings are retained. 6.11 This Specification is included in the LTDTS as Specification D6.11. Reference to 10 CFR 19 is added to this Specification. 6.12 This ' Specification was previously deleted. 6.13 The requirements for access control to High Radiation Areas is included in the LTDTS as Specification D6.12. 6.14 This Specification, Environmental Qualification, is not applicable in the LTDM since an accident that results in a harsh environment which could affect the ability to maintain SFP Water level or temperature is not credible in the LTDM; therefore, this Specification is not included in the LTDTS. See the evaluation for Specification 6.10.21 on page 32 of this SAR. 6.15 This Specification, PCP, is revised in accordance with the guidelines of Generic Letter 89-01, and is included in the LTDTS as Specification 06.13.

( ' License Change Safety Analysis Log No. 1091 Revision 1 Proposed Amendment _No. 182, Revision.1 Page 33 of 38 Technical Specification Number Discussion 6.16 This Specification, ODCM and REHP, is revised in accordance with - the guidelines of Generic Letter 89-01, and is included in the LTDTS as Specification D6.14. 6.17 This Specification, Major Changes to Radioactive Waste 1reatment Systems, is celocated to the PCP in accordance with the guidelines-of Generic Letter 89-01. Major changes to radioactive waste treatment systems will be reviewed by the NSRAC as required by LTDTS Specification D6.5.1.71. 6.18 This Specification, Postaccident Sampling (PAS),'is not applicable in the LTDM because the PAS system is designed to monitor reactor accidents, and thus this Specification is not included in the LTDTS. The requirements of Appendix B.to Operating License No. DPR-54 have been previously justified for deletion from the Technical Specifications in Proposed Amendment No. 102. The administrative controls that were required to y i be retained in Appendix A so that the Appendix B Specifications could be deleted are contained within the LTDTS (Appendix C). Sufficient justification l: and administrative controls exist in this submittal and previous submittals (PA-102)- to conclude that the Appendix B Specifications are adequately addressed in the LTDTS for the LTDM. 1

F . License Change Safety Analysis Log N3. 1091, Revisitn 1 Proposed Amendment No.182, Revision 1 Page 34 of 38 TABLE 1-TIME TO REACH 212*F FROM 120'F-TIME TO REACH AND VAPORIZE SPENT FUEL 212*F FROM 6.75 FT. OF DECAY HEAT 120*F IN THE SPENT FUEL (MILLION SPENT FUEL POOL POOL HATER QAIE BTU /HR) (HOURS) (DAYS) (HOURS) (DAYS) 1/3/90-3.600 74.4 3.1 220.25 9.18 6/7/90-2.786 96.12 4.01 284,56 11.86 6/7/91 1.862 143.85 5.99 425.86 17.74 6/7/94 1.372 195.21 8.13 577.88 24.08 6/7/99 1.182 230.97 9.62 683.76 28.49 6/7/09-0.897 298.73 12.45 884.33 36.85 n

' License Chang'; Safety Analysis - Log N6. 1091, Revisicn 1-Proposed' Amendment No.182,. Revision 1 Page 35 of 38 l

SUMMARY

l l The Rancho Seco LTDTS provide the controls necessary for the protection of the I . health and safety of the public in the LTDM. These LTDTS are intended to act l-as a stand alone document and to take the-place of Appendices.A and B to j Operating License No. DPR-54 during the current extended outage while the j plant is in the new defined mode LTDM. The LTDTS were developed based on an evaluation of credible accidents -applicable ~.in the LTDM,:and the equipment needed to mitigate these accidents. l Two accidents evaluated in Chapter 14 of the USAR are considered credible in l the LTDM; 1) a fuel handling accident, and 2) a LOOP. The LOOP accident is 1 p determined to have no adverse-impact on the health and safety of the public in t 1 the LTDM, while the dropped fuel assembly accident results in very small calculated exposures to the maximum exposed individual. The existing Rancho Seco Technical Specifications (Appendices A and B) were evaluated and the appropriate requirements incorporated into the LTDTS and other plant documents (i.e., ODCH, REMP Manual, and PCP) to ensure that plant closure activities will result in radioactive releases from the plant which are as low.as is reasonably achievable. A detailed description of this . evaluation is.provided in this safety analysis.- The LTDTS presented in Proposed Amendment No.182, Revision 1 provide . technical and administrative controls. sufficient to assure the-protection of the. health and safety of the public. For the plant in the LTDM, the LTDTS (Appendix C) maintain the margin of safety previously provided by the Appendix. A and B Technical Specifications when the plant was operating. l i l

n i-a License Change Safety Analysis Log ND,1091, Revisitn 1 Proposed Amendment No.182, Revision 1 Page 36 of 38 N0'SIGNIFICANT HAZARDS CONSIDERATION The District has reviewed the proposed LTDTS against each of the criterion of 10 CFR 50.92 and concluded that the proposed changes as described and evaluated in the above safety analysis do not: a. Involve a significant increase in the probability or consequences of an accident previously evaluated. There are only two credible accidents in the Defueled Mode, a fuel handling accident and a LOOP. The changes proposed do not increase the probability of either of these accidents since the LOOP is not controllable by the plant, and the requirements for testing of the fuel handling bridge remain unchanged. The consequences of the two credible accidents are bounded by the previous analyses for these accidents. The fuel handling accident scenario remains unchanged, but the consequences of this accident is reduced due to the length of the decay time. The extended time period available to restore offsite power before only 6.75 feet of water boils from the SFP (a minimum of 9.18 days as of January 3,1990) provides ample time to take corrective action to ensure fuel cladding damage does not occur following a loss of offsite power.- b. Create tne possibility of a new or different kind of accident from any accident previously evaluated. The proposed addition to the Operating License, the LTOTS (Appendix C), applies only when the reactor is in the LTDM. Therefore, only those activities and potential accidents associated with the SFP need be considered. Because the ability to maintain an adequate shutdown margin in the SFP is not affected, no physical changes-l to the SFP or SFC are being made, and systems required to safely store the i spent fuel in the LTDM will continue to be maintained, the possibility of a new or d fferent kind of accident-from any accident previously evaluated is not created, i c. Involve a significant reduction in a margin of safety since the margin of safety for the fuel handling accident is unchanged, while the margin of l safety for the :.00P event is not significantly reduced given the extended time period available to restore offsite power or to find an alternate means of adding water to the SFP. 1 Based on the evaluation provided above, the District has concluded that the proposed LTDTS do not constitute a significant hazard to the public and do not endanger the public's health and safety. u-m--

License Change Safety Analysis Log No.1091, Revision 1 j - Proposed Amendment No.182. Revision 1 Page 37 of 38 l i i REFERENCES ~ j 1.. Proposed Amendment No. 182 dated December 28, 1989, addition of Appendix C to Operating License No. DPR-54. This submittal contains the ODCM, REMP Manual, and PCP to support the { implementation of Generic Letter 89-01. 2. Generic letter 89-01, IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS IN THE ADMINISTRATIVE CONTROLS SECTION OF THE TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF RETS TO THE OFFSITE i DOSE CALCULATION MANUAL OR THE PROCESS CONTROL PROGRAM ^; 3. Generic Letter 89-14, LINE-ITEM IMPROVEMENTS IN TECHNICAL SPECIFICATIONS - REMOVAL OF THE 3.25 LIttIT ON EXTENDING SURVEILLANCE INTERVALS i

4. -

Generic Letter 88-12, REMOVAL OF FIRE PROTECTION REQUIREMENTS FROM ~ TECHNICAL SPECIFICATIONS. 5. Proposed Amendment No. 102 dated December 12, 1984, removal of Appendix B from Operating License No. DPR-54. 6. Proposed Amendment No.102, Revision 1 dated July 27, 1988, -removal of Appendix B from Operating License No. DPR-54. 7. Proposed Amendment No. 102, Revision 1, Resubmittal dated July 5, '1989, removal of Appendix B from Operating' License No. DPR-54. 8. SMUD calculation No. Z-RCS-N0046, Decay Heat Power After 6/7/89 - Cycle 7 9. SMUD calculation No. Z-SFC-N0046, Spent Fuel Heat Generation Following June 7, 1989 Shutdown Of Rancho Seco 10. SMUD Calculation No. Z-SFC-M2533, Time To Boil And Time To Boil Down The Fuel Pool 11. SMUD Calculation No. Z-SFC-N0049, Maximum Predicted Whole Body And Child Thyroid Dose Rates At The Site Boundary From Postulated Accidents During Plant Shutdown 12. SMUD Calculation Z-SFC-0029 Spent Fuel Pool - Surface Dose Rate

1 t: t License Change Safety Analysis Log No.1091, Revisitn 1 Proposed Amendment N3. 182, Revisicn 1 Page 38 cf 38 u REFERENCES (Continued) 13. NRC to SMUD letter dated September 13, 1982, TECHNICAL SPECIFICATIONS (TSs) FOR EMERGENCY SAFETY FEATURE (ESF) AIR i FILTERS, Safety Evaluation Report for Amendment No. 39 to Facility b Operating License No. DPR-54 14. SMUD to NRC letter dated April 17, 1989, 10CFR50.63, LOSS OF ALL ALTERNATING CtJRRENT POWER 15. Regulatory Guide 1.155, STATION BLACK 0UT, June 1988 16. Rancho Seco Nuclear Generating Station, Unit No.-1, Updated Safety l Analysis Report, Amendment 6 17. NUREG-0103, Rev.4, Standard Technical Specification l 18. Proposed Amendment No. 180 dated December 28, 1989, removal of the i Fire Protection Requirements from the Technical Specifications per l Generic Letter 88-12. - 19. Proposed Amendment No. 180, supplement 1 dated March 16, 1990, response to NRC comments. l l i i + t

4 F,'<. i ), h,' t ~! r 4 \\ b:d t, l ATTACHMENT III 1 CERTIFIED FUEL HANDLER TRAINING PROGRAMS t r l. I 1' l t i l j L. e.4}}